Stichting Laka: Documentatie- en onderzoekscentrum kernenergie De Laka-bibliotheek Dit is een pdf van één van de publicaties in de bibliotheek van Stichting Laka, het in Amsterdam gevestigde documentatie- en onderzoekscentrum kernenergie. Laka heeft een bibliotheek met ongeveer 8000 boeken (waarvan een gedeelte dus ook als pdf), duizenden kranten- en tijdschriften- artikelen, honderden tijdschriftentitels, posters, video’s en ander beeldmateriaal. Laka digitaliseert (oude) tijdschriften en boeken uit de internationale antikernenergie- beweging. De catalogus van de Laka-bibliotheek staat op onze site. De collectie bevat een grote verzameling gedigitaliseerde tijdschriften uit de Nederlandse antikernenergie-beweging en een verzameling video's . Laka speelt met oa. haar informatie- voorziening een belangrijke rol in de Nederlandse anti-kernenergiebeweging. The Laka-library This is a PDF from one of the publications from the library of the Laka Foundation; the Amsterdam-based documentation and research centre on nuclear energy. The Laka library consists of about 8,000 books (of which a part is available as PDF), thousands of newspaper clippings, hundreds of magazines, posters, video's and other material. Laka digitizes books and magazines from the international movement against nuclear power. The catalogue of the Laka-library can be found at our website. The collection also contains a large number of digitized magazines from the Dutch anti-nuclear power movement and a video-section . Laka plays with, amongst others things, its information services, an important role in the Dutch anti-nuclear movement. Appreciate our work? Feel free to make a small donation . Thank you. www.laka.org | [email protected]| Ketelhuisplein 43, 1054 RD Amsterdam | 020-6168294
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Stichting Laka: Documentatie- en onderzoekscentrum kernenergie
De Laka-bibliotheek
Dit is een pdf van één van de publicaties in de bibliotheek van Stichting Laka, het in Amsterdam gevestigde documentatie- en onderzoekscentrum kernenergie.
Laka heeft een bibliotheek met ongeveer 8000 boeken (waarvan een gedeelte dus ook als pdf), duizenden kranten- en tijdschriften-artikelen, honderden tijdschriftentitels, posters, video’s en ander beeldmateriaal. Laka digitaliseert (oude) tijdschriften en boeken uit de internationale antikernenergie-beweging.
De catalogus van de Laka-bibliotheek staat op onze site. De collectie bevat een grote verzameling gedigitaliseerde tijdschriften uit de Nederlandse antikernenergie-beweging en een verzameling video's.
Laka speelt met oa. haar informatie-voorziening een belangrijke rol in de Nederlandse anti-kernenergiebeweging.
The Laka-library
This is a PDF from one of the publications from the library of the Laka Foundation; the Amsterdam-based documentation and research centre on nuclear energy.
The Laka library consists of about 8,000 books (of which a part is available as PDF), thousands of newspaper clippings, hundreds of magazines, posters, video's and other material. Laka digitizes books and magazines from the international movement against nuclear power.
The catalogue of the Laka-library can be found at our website. The collection also contains a large number of digitized magazines from the Dutch anti-nuclear power movement and a video-section.
Laka plays with, amongst others things, its information services, an important role in the Dutch anti-nuclear movement.
Appreciate our work? Feel free to make a small donation. Thank you.
electricity sales at competitive prices with other sources apart from any subsidies
on those is required (and without any subsidy beyond what is required to counter
market distortions due to those sources).
But the clear message from practically every international authority and their
reports is that nuclear power is essential for meeting the world’s growing need for
affordable, clean, and reliable electricity. There is no credible reason to not greatly
increase its role in world electricity production, and for industrial heat including
desalination.
1.2 Energy density, other characteristics
The single most remarkable characteristic of uranium is its energy density: the amount
of energy that a single kilogram can yield. Even so, exactly how much energy depends
on the technology used to liberate it. One fuel pellet the size of a fingertip produces as
much energy as one tonne of coal, even in the least efficient reactor.
“Natural uranium” is that which is found in nature—mostly in the Earth’s crust
in a variety of geological environments. In most nuclear reactors, only about half of
1% of this is actually used, but even so it yields about 500 GJ/kg, about 20,000
times as much as black coal. In a fast neutron reactor, about 60 times this is achiev-
able, and one day potentially more if technology and economics were pressed. But,
in fact, uranium is fairly common and not a high-priced commodity—it has been
less than $100/kg in recent years—so there is little incentive as of yet to push those
boundaries from a resource perspective.
Like most other elements, uranium occurs as a mixture of isotopes, but with ura-
nium, that fact is central to its use. Only one of the natural isotopes is directly usable,
and that comprises only 0.7% of natural uranium. Hence, either power plants need to
be designed accordingly, or the uranium needs to be enriched in that minor isotope,
which is the subject of Chapters 11 and 12. In fact, the latter course of action
accounts for 88% of the world’s nuclear power reactors (and all naval reactors).
Also, like most metals, uranium occurs in a variety of chemical forms, though
most is as a mixed oxide of UO2 and UO3, characteristically U3O8. Chapter 2 offers
a deeper discussion.
The concentration of uranium in its geological settings can range up to about
20% in some Canadian deposits, though 2%U is generally called a high-grade ore.
Geological occurrences where it is defined as “ore” (economically recoverable)
range down to about 0.01%U.
1.3 Resource situation
Uranium is approximately as common in the Earth’s crust as tin and zinc, and
occurs in most rocks. Granites typically have up to 5 ppm U (which incidentally
and at higher levels provides the heat for geothermal energy). Seawater contains a
vast amount at 0.003 ppm U, which is recoverable, but not economical.
4 Uranium for Nuclear Power
Our knowledge of what uranium (or anything else) is in the Earth’s crust arises from
mineral exploration activities (see Chapter 3), which are expensive and mostly under-
taken by mining companies that have negotiated the right to mine what they find and
quantify. Therefore, any figures published based on this refer only to known resources.
Beyond geological theorizing, we have little idea of what there is beyond this.
The world’s known resources related to cumulative exploration expenditure are
shown in figure.
It can be seen how increased exploration leads to increased known resources. In
2013, there were 5.9 million tonnes of uranium known and recoverable at up to
$130/kgU, and a bit more at double that. The graph shows only historical figures,
without allowing for inflation. Chapter 4 provides more detail, and Chapter 5 looks
at the complexities of the uranium market.
1.4 Technological perspective
Mention has been made of reactor technologies to liberate uranium’s energy in a
controlled fashion. Prior to discussing these, it is necessary to outline a little basic
physics. The objective in any nuclear reactor is to achieve criticality, which results
in an ongoing chain reaction of nuclear fission, essentially of the uranium-235
(235U) isotope’s nucleus. This means that each 235U atomic nucleus splits into two
5Uranium for nuclear power: an introduction
parts (fission products) and releases two or three neutrons in the process. These are
“fast” neutrons with high energy. But to keep the chain reaction going, with one
nuclear fission producing one more nuclear fission, the neutrons need to be slowed
down to about one-ten thousandth of their velocity, by a moderator.
The most efficient moderators are graphite and heavy water. With a reactor
employing these, natural uranium can be used as fuel. However, for cost and techni-
cal reasons, these are not usually preferred, and the main moderator used (in 88%
of the world’s power reactors and all naval ones) is ordinary “light” water, which
doubles as coolant. But this means that the fuel needs to be enriched to at least 3%235U, and typically today is enriched to 4�5%. (Naval reactors use much higher
enrichment to give a more compact unit.) So uranium enrichment plants are a major
feature of the world’s nuclear fuel cycle, as will be explained in Chapter 12.
It is also possible to have a reactor running with fast neutrons, though these are a
little more complex, and bring plutonium more fully into the picture. Most generation
IV nuclear reactors expected to be deployed from about 2030 are fast reactors.
Although a normal nuclear reactor is usually loaded with only uranium fuel, in
fact about one-third of its energy output comes from plutonium. How so? Well,
99% of natural uranium is 238U, which is said to be “fertile” rather than “fissile,”
and in a reactor, it captures some of those spare neutrons to become plutonium,
principally 239Pu. 239Pu fissions in the same way as 235U.
In a fast neutron reactor, plutonium has a greater role, and there is no moderator.
Cooling the core requires a heat transfer fluid with minimal moderation, hence
liquid metals such as sodium are used, and perhaps fluoride salts in the future.
To generate electricity, the physics described is applied in a plant designed to
achieve high temperatures in the core to make steam that drive turbines. A coolant,
most commonly water under considerable pressure, transfers the heat from fission
in the core at a little over 300 �C to make steam either above the core or in separate
steam generators. Over 400 of such reactors are operable for power generation, and
another 180 or so smaller ones are used for naval propulsion.
About 240 “research” reactors are designed simply as neutron factories, and
these are much smaller and operate at much lower temperatures. The neutrons are
beamed out from the core for research, production of medical and industrial
isotopes, or other purposes.
Since the 1940s, thorium has been considered as a possible source of nuclear
fuel. Although it has no fissile isotope, it does produce fissile 233U by neutron
capture in a nuclear reactor, and this is an attractive source of energy, like 235U
and 239Pu. However, some technical challenges remain for the time being and
Chapter 10 describes the situation.
1.5 The electromobility frontier and methanol
While France has shown the world that well over half of the electricity for a major
industrial country can economically be generated by nuclear power, a high
6 Uranium for Nuclear Power
proportion of electricity from nuclear plants in a system creates the need for
either load-following to reduce output overnight, or some other application
to compensate for temporary low system demand (in the absence of large-scale
storage).
Increasing world interest in battery-powered cars and light commercial vehicles
is helpful here, if batteries can be charged off-peak and overnight. This will have
the effect of increasing the proportion of total electricity that can be generated by
base-load plants, that is, on a continuous 24/7 basis, which will in most circum-
stances reduce the unit (per kilowatt-hour) cost of all electricity. If those base-load
plants are nuclear rather than coal-fired, then automotive as well as the electricity
usage is virtually emission-free.
Estimates vary concerning the potential effect of this electromobility on electric-
ity demand, but a 20% increase seems a conservative estimate if there is widespread
change from petrol/gasoline and diesel power to electric.
Another way nuclear power may become important for motor vehicles is
making hydrogen which is used for converting carbon dioxide into methanol. This
has about two thirds the energy density of gasoline and utilises present internal
combustion engines. Methanol can be dehydrated to dimethyl ether (DME), a
good diesel fuel.
1.6 Relationship with nonhydro renewables
Government policies in many countries demand that nonhydro renewables, mostly
wind and solar power, should supply a significant proportion of electricity.
Economic incentives are applied to achieve this, regardless of actual generating
costs from those sources, and regardless of how well the supply from those sources
over any 24 h may fit the normal demand pattern that it is vital for transmission
operators to fully and reliably maintain supply.
The economic characteristics of any generating plant with high capital costs and
low running costs—such as nuclear—mean that it is best run continuously, leaving
other (eg, gas-fired) plants with lower capital costs and higher running costs to
operate for only part of each day, picking up peak loads. If a significant amount of
electricity from intermittent and unpredictable wind and solar sources is superim-
posed on this, so that output from other plants is curtailed, then costs of power from
those plants increase. If this cost increase is added to the subsidies and other sup-
port for those disruptive renewables, then the effect on the ultimate consumer can
be great. Of particular relevance here is that the economic virtues of nuclear power
are not properly exploited.
But apart from economics, any major input of intermittent and unpredictable
renewables means that major challenges arise for transmission operators to keep
the grid stable and reliable, a role for which nuclear power has long been
appreciated.
7Uranium for nuclear power: an introduction
1.7 Safety, regulation
From the outset in mid-20th century, the energy density of nuclear power has made
safety a high priority. This is achieved by engineering, both in the quality of
containment and the physics of the reactor systems.1 A considerable part of the
engineering in any nuclear power plant is back-up provision in design and safety
equipment, which has only ever been fully called upon twice in the history of
nuclear power.2
To achieve optimum safety, nuclear plants3 operate using a “defense-in-depth”
approach, with multiple safety systems supplementing the natural features
of the reactor core. Key aspects of the defense-in-depth approach include the
following.
● High-quality design and construction● Equipment that prevents operational disturbances or human failures and errors developing
into problems● Comprehensive monitoring and regular testing to detect equipment or operator failures● Redundant and diverse systems to control damage to the fuel and prevent significant
radioactive releases● Provision to confine the effects of severe fuel damage (or any other problem) to the
plant itself
These steps can be summed up as prevention, monitoring, and action (to mitigate
consequences of failures).
Looked at functionally, the three basic safety functions in a nuclear reactor are
to control reactivity, to cool the fuel, and to contain radioactive substances.
The main safety features of most reactors are inherent—negative temperature
coefficient and negative void coefficient. The first means that beyond an optimal
level, as the temperature increases, the efficiency of the reaction decreases (this in
fact is used to control power levels in some new designs). The second means that if
any steam has formed in the cooling water, there is a decrease in moderating effect
so that fewer neutrons are able to cause fission and the reaction slows down
automatically.
Regulation of nuclear facilities is undertaken nationally, but with a great deal of
international collaboration. Reactor designs are subject to close scrutiny, as is
construction and then operation.
Nuclear power is safe, with the best record of any major form of electricity
generation.
1 For instance, if water boils in the core of a light water reactor, moderation is reduced and the chain reac-
tion slows. Or in a fast or high-temperature reactor, very high core temperatures slow the chain reaction.2These two times are at Three Mile Island in 1979 and Fukushima Daiichi in 2011. No one was killed or
seriously harmed in either accident. Chernobyl had a less-than-standard safety provision, and perhaps
50 deaths were attributed to the accident directly and indirectly.3Once one might have added “in the Western world,” but today high standards prevail in both the east
and west.
8 Uranium for Nuclear Power
1.8 Nonpower uses
Apart from charging batteries for cars and light commercial vehicles referred to
previously, which have a potentially significant effect on how nuclear or other
base-load provisions fit into the whole electricity demand picture, there are a
number of other uses for nuclear energy fueled by uranium.
The most obvious of these is desalination of seawater or brackish groundwater to
produce potable water for cities. This may use either electricity (ideally off-peak) or
waste heat from electricity generation. Most desalination today uses fossil fuels, and
thus contributes to increased levels of greenhouse gases. Total world capacity in
2013 was 80 million m3/day (29,200 GL/year) of potable water in over 17,000
plants. A majority of these are in the Middle East and north Africa. Two-thirds of the
world capacity is processing seawater, and one-third uses brackish artesian water.
Direct industrial uses of nuclear heat are also possible. The potential application
of nuclear heat depends mainly on the temperature required. Most of today’s reac-
tors produce heat at a little over 300 �C; fast reactors go to 550 �C. With reactor
output temperatures of up to 700 �C, there is a wide range of possible applications:
at 900 �C, there are further possibilities, and at 950 �C, an important future applica-
tion to hydrogen production opens up. About 20% of US energy consumption
goes into process heat applications, compared with 35�40% into electricity. In this
15�20%, replacing fossil fuels with nuclear heat promises much in energy security,
price stability, and reduced emissions.
1.9 Wastes, radiation, proliferation
No source of energy is without issues that, if not managed properly, are negatives.
Coal has been a massive benefit to humanity since the Industrial Revolution, but
has killed many in mine accidents, occupational health effects, and environmental
pollution. Today, with those issues mostly under control, it is in disrepute because
of its CO2 emissions.
Nuclear power has at least the same potential as coal to benefit humanity, but
only if its wastes continue to be properly contained and managed, if ionizing radia-
tion arising directly and indirectly from it causes no harm, and if it clearly does not
increase the risk of nuclear weapons proliferation.
Waste from nuclear power, considering the whole fuel cycle from uranium min-
ing onward, range from innocuous to very hazardous. They are managed and if nec-
essary contained accordingly. The high-level wastes comprising either used fuel or
the fission products and actinides separated from it, need both cooling and shield-
ing. This is straightforward, and has been done for more than half a century with no
harm to anyone and no environmental effects. Long-term disposal of this waste is
also technically straightforward—deep geological disposal being the universal
choice once the radioactivity has decayed to less than 1% of its original level.
Ionizing radiation at low levels is ubiquitous and experienced by all of us
constantly, and the nuclear industry, under the watchful eye of regulators, ensures that
does not add significantly to natural background levels for neighbors, or reach more
9Uranium for nuclear power: an introduction
than a fraction of harmful levels for workers. The Chernobyl accident is the one occa-
sion when workers (and firefighters) received harmful radiation doses. (In contrast,
radioactive sources and accidents outside the civil nuclear industry have led to signifi-
cant harm on a few occasions.) Chapter 16 addresses occupational health aspects of it.
Civil nuclear power arose by new application of the technology developed for
bombs in World War II. The military links continue, but all one way: in every
country that has both nuclear weapons and nuclear power, the weapons came first,
and more recently, substantial amounts of uranium in military stockpiles have been
diluted and used for electricity generation—about 10% of US electricity from 1993
to 2013 came from Russian military stockpiles. See also Chapter 8.
But at the level of routine uranium supply, international accounting and auditing
procedures called safeguards are applied to ensure that any diversion from civil to
military application is detected and high-level diplomatic pressure can be applied as
a deterrent. Similar provisions apply regarding the use of equipment, and recently
Iran’s uranium enrichment program has been the subject of sanctions applied by the
UN Security Council. Safeguards are applied under the Nuclear Non-Proliferation
Treaty of 1970, to which practically all countries subscribe.4
Nuclear technology for making weapons is in the public domain, and no country
is without enough uranium to make a few bombs. The Nuclear Non-Proliferation
Treaty provides strong disincentives against weapons proliferation by any country
beyond the nine that now have nuclear weapons.
1.10 Uranium in the future
There are a number of authoritative projections from the Organisation for
Economic Co-operation and Development (OECD) and UN sources concerning the
future of energy supplies in general and electricity in particular. Practically all of
these show a considerably increased role for nuclear energy in meeting human
needs much more widely than today. Therefore, uranium will remain a vital energy
source, even if by mid-century much of it is essentially recycled, drawn from the
huge stockpiles of depleted uranium arising from a century’s enrichment activity.
The assembled contributions comprising this book set the scene for a much
greater role for uranium in the decades ahead. Readers are urged to look beyond
these pages to what is happening in more than 30 countries that are implementing
plans and making heavy investments in nuclear power plants that depend on the
world’s abundant uranium.
1.11 Further information
World Nuclear Association: Information Library http://www.world-nuclear.org/
Information-Library/
4The exceptions are India, Pakistan, Israel, and North Korea. Since 2008, India has effectively come
under safeguards. Pakistan has its civil nuclear facilities under safeguards. Israel and North Korea have
Basin and Range (USA-Mexico) Volcanic-related 14.000
Singhbhum District (India) Metamorphite 70.000 11.000
Hercynian�Andean Province (Peru, Bolivia,
Argentina)
Granite-related
Metamorphite
Sandstone
Surficial
n.a.5 non available
several percent Ca, Si, and/or Zr, and is named pitchblende. When occurring as pulveru-
lant microcrystals, this mineral is named “sooty pitchblende.” Coffinite, USiO4, is the sec-
ond most common mineral in uranium ore deposits and occurs either as a primary ore
mineral mostly in sandstone-hosted deposits or as a common alteration product of urani-
nite in most deposits. Uranium may also occur as ningyoite (U,Ca,Ce)2(PO4)2,1.5H2O) in
some rollfront-type deposits and brannerite (U,Ca,Ce)(Ti,Fe)2O6 mainly in hydrothermal
metasomatic deposits. More rarely, uranium occurs in uranothorianite (Th,U)O2 in skarn
deposits and uranothorite (Th,U)SiO4 in deposits associated with peralkaline magmatism.● Nb-Ta-Ti oxides form a wide family of minerals with the predominance of one or two of
the cations Nb, Ta, and Ti and the presence of other cations such as REE0, Y, and Ca.
Their uranium content is quite variable but may reach up to 18 wt% UO2. The main
minerals are uranmicrolite (U,Ca,Ce)2(Ta,Nb)2O6(OH,F), uranpyrochlore (U,Ca,Ce)2(Nb,
Ta)2O6(OH,F), betafite ((Ca,Na,U)2(Ti,Nb,Ta)2O6(OH,F)), and euxenite (Y,Ce,La,U)(Nb,
Ti,Ta)2(O,OH)6. They are commonly present in uranium deposits associated with peralka-
line magmatism, but betafite may occur in intrusive anatectic deposits, as at Rossing.
They represent refractory uranium sources and the cost of uranium extraction during ore
processing is high.● Hexavalent uranium minerals are highly colored and some fluoresce under ultraviolet
light. More than 200 species exist. They can be deposited either as primary ore minerals
such as carnotite K2(UO2)2 (VO4)2,3H2O in calcrete-type deposits, but more commonly,
hexavalent uranium minerals represent alteration products of uranium oxides by oxidizing
surface water or groundwater, mostly in the uppermost parts of deposits (Finch and
Murakami, 1999) such as autunite Ca(UO2)2(PO4)2,10H2O and uranophane Ca(UO2)2SiO3(OH)2,5H2O. For this reason, they are also commonly referred to as secondary ura-
nium minerals.
Accessory minerals. These minerals occur in low abundance in most rocks and
incorporate trivalent elements such as REE0 and Y; quadravalent elements such as
U, Th, Zr, and Hf; and pentavalent elements such as Nb and Ta. They generally
contain a few tens to several thousands ppm U, rarely up to several percent. They
are highly concentrated in sedimentary placers and in fractionated peralkaline igne-
ous rocks. A complete solid solution exists with Th41 in uranothorianite and ura-
nothorite may contain up to 30% UO2 (Cuney and Friedrich, 1987). More limited
substitution of U41 occurs for Ca21 in fluorapatite (Ca)5(SiO4)3(F,OH) and allanite
(Ce,Ca,Y,Th,U)2(Al,Fe,Mg)3(SiO4)3(OH), for Zr41 (0.84 Ǻ) in zircon (Zr,U)SiO4,
REE in rare earth fluorocarbonates such as bastnaesite, (LREE,U)CO3F, and in
phosphates such as monazite (LREE, Th,U,Ca)[PO4, SiO4], and xenotime (Y,
HREE,U)PO4.
Adsorbtion. The adsorption capacity of uranium by some mineralsmay in some
cases dominantly control its geochemistry, most particularly in the exogenous cycle.
Iron, titanium oxides, and oxi-hydroxides are the most effective adsorbents of ura-
nium under near-neutral pH conditions. The adsorption capacity of clayminerals for
uranium is lower and varies with the surface charge increasing from kaolinite, to
illite, to montmorillonite, and to interlayered clay minerals.
Associated with organic matter. The spatial connection between uranium and
organic matter is often observed mostly in sedimentary but also in other rock types.
It is related to the power of complexation and reduction of organic matter.
16 Uranium for Nuclear Power
Uraniumin major rock forming minerals. This is only present at the ppb level
because of its large ionic radius and high valency. When detected in higher concen-
trations, this reflects the presence of microinclusions of uranium-bearing accessory
minerals or adsorption along cleavage planes or fractures.
Dissolved in geologic fluids and fluid inclusions. In most geologic fluids, ura-
nium concentration is below ppb levels and reaches 3.3 ppb in seawater. Meteoric
water in equilibrium with uranium oxides may reach 1�2 ppm U. The highest ura-
nium concentrations, reaching several hundred ppm, have been measured in fluid
inclusions corresponding to magmatic fluids and to fluids associated with the gene-
sis of unconformity uranium deposits.
2.4 Classification of uranium deposits
Various global and regional classifications of uranium deposits have been published
previously. These classifications generally follow two alternative approaches, focus-
ing either on descriptive features of the mineralization, such as host rock type and
orebody morphology (geological classification), or on genetic aspects (genetic clas-
sification). Other classifications considering the geotectonic position of deposits
(tectono-lithologic classification) have also been published, in particular by Russian
geologists for various regions of the world.
2.4.1 The IAEA geological classification of uranium deposits
In 2010, a working group was created by the IAEA to review the various existing
classifications and propose a new or a modified classification that would be adopted
internationally. The IAEA classification, used in particular in the 2012 Red Book
(OECD/NEA-IAEA, 2012), dated back to 1993 and increasing availability of infor-
mation about the deposits from Eastern countries, used extensively by Dahlkamp to
establish his new classification (Dahlkamp, 2009), new publications and a great
increase in company data following the increase of uranium prices starting in 2005
all provided a wealth of information on uranium deposit geology that was used to
revise the classification.
The revised classification of uranium deposits was officially accepted in 2013
(Bruneton et al., 2014) and is presented in an IAEA report “Geological classifica-
tion of uranium deposits and description of selected examples” with publication
planned for 2016 (IAEA, 2016a). The revised classification is used in the 2014 ver-
sion of the Red Book (OECD/NEA-IAEA, 2014).
Fifteen types of deposits have been retained in the new IAEA classification
scheme. They are listed in order from deep primary magmatic deposits to sedimen-
tary and surficial deposits following the geological cycle of Cuney (2010) illus-
trated in Fig. 2.1. The economic ranking used in the previous IAEA classifications
has not been taken into account.
17Geology of uranium deposits
Most subtypes and classes defined by Dahlkamp (2009) have been retained with
minor modifications and some new subtypes and classes have been created. Most
deposit types are named by the host rocks except for types 3, 7, and 8 (which are
related to structures) and type 5 (which is related to metasomatic alteration).
2.4.1.1 Detailed geological classification with types, subtypes,and classes
Guinea Firawa 7.890 0.0283 Metamorphite Structure-bound Forte Energy
Namibia Aussinasis 6.960 0.02 Intrusive Anatectic Deep Yellow Ltd
Turkey Temrezli 6.700 0.099 Sandstone Tabular Anatolia Energy Corp
Canada GemicoBlock 3 6.262 0.017 QPC U-dominant Appia Energy Corp
Canada GemicoBlock 10 5.977 0.039 QPC U-dominant Appia Energy Corp
excess of 900.000 tU, the Namib Desert is one of the largest uranium provinces in
the world.
Type 2—Granite-related: Exploration in historical districts has taken place in
Spain and China districts. In Spain, new resources have been defined in several
perigranitic deposits and prospects. The district contains about 44.000 tU within 20
deposits.
Type 4—Volcanic-related: Several discoveries have been made in the volcanic
Macusani District (Peru) such as Kihitian. At Macusani, the volcanic-related depos-
its are hosted in felsic peraluminous pyroclastic rocks. The district has 41.000 tU of
low-grade geological resources in 13 deposits.
Types 5—Metasomatite: No important discoveries were related to this type of
deposit but worthmentioning is the discovery of Inca (Namibia), an unusual skarn
subtype deposit associated with the alaskites.
Types 6—Metamorphite: Only one deposit of this type, J4�Ray, was discovered
in Nunavut (Canada), near the Lac Cinquante Deposit, whose resources have been
significantly increased.
Type 7—Proterozoic unconformity: Several Proterozoic unconformity deposits
were discovered in the Athabasca Basin (Canada) and surrounding areas, such as
Triple R, Phoenix, Roughrider, and J-Zone. The Triple R, the most recent discovery,
is a large, basement-hosted deposit situated outside the Athabasca Basin, along its
southwestern margin. The Athabasca Basin has resources of 880.000 tU within 55
deposits. In Australia, the Ranger Deep deposit situated below the mined-out Ranger
3 deposit was confirmed and is presently being developed for underground mining.
Type 9—Sandstone: Important resources have been discovered or confirmed in
the Tim Mersoi Basin (Niger). Daja and Dajy are new discoveries and Marianne-
Marylin, Isakanan, and Abakorum hadresources confirmed or increased. Total geo-
logical resources of the Tin Mersoi Basin are 740.000 tU in 42 sandstone-tabular
deposits.
The Frome Embayment (Australia) confirmed its potential with the discovery of
Four Mile East and Four Mile West. Geological resources of the district near the
Beverley mine are 68.000 tU in seven sandstone (basal channel, roll and tabular)
deposits.
Two low-grade deposits in sandstone were discovered in the Sainshand Region
of Mongolia (Zoovch Ovoo and Dulaan Ull) and two in the Erlian Basin, China
(Bayinwula and Sunjialang). They are rollfront deposits in Mesozoic formations.
The Zoovch Ovoo deposit has resources in excess of 50.000 tU.
In Kazakhstan, resources have been greatly increased and confirmed in known
deposits. Geological resources are on the order of 850.000 t in about 30 deposits,
some of them with resources larger than 100.000 tU (Inkai, Mynkuduk). They all
correspond to sandstone deposits associated with stacked rollfronts.
The Karoo Formation in east Africa, running from South Africa to Tanzania
through Zambia and Malawi, was the focus of exploration for numerous companies.
Today, resources stand at 260.000 tU located within 25 deposits. They are all low-
grade sandstone-tabular deposits with the largest one Nyota (Tanzania) containing
55.000 tU in several lenses.
43Geology of uranium deposits
Type 10—Paleo-QPC: Exploration drilling has restarted in the historical Elliott
Lake District, not only for U, but also for REE. Seven explored deposits contain
geological resources of 101.000 tU. The two largest ones are Banana Lake and the
Eco Ridge Mine.
Type 11—Surficial: Two important new surficial deposit districts were delin-
eated on the Reguibat Shield of Mauritania (19.400 tU) and the Manyoni District
(11.450 tU) in Tanzania. Several deposits were discovered and/or confirmed in
Western Australia, which contains 100.000 tU in 18 deposits and in Namibia with
resources of more than 205.000 t in 15 deposits.
Type 13—Carbonate: A large unusual strata-bound deposit, Tummalappalle,
has been defined in the dolomitic Vempalle Formation of Proterozoic age
(India). Resources are on the order of 72.000 tU and the deposit is open in all
directions.
2.7.2 Unconventional deposits and resources (Table 2.10)
Type 1-Subtype 1.2—Intrusive plutonic (quartz monzonite, peralkaline complex, and
carbonatites):Resources were greatly increased at the Kvanefjeld REE-U-Zn project
with 228.240 tU within three deposits. In Morocco, resources have been published
for two carbonatite intrusions and one peralkaline complex. Twihinate, with
118.500 tU, is the largest.
Type 3—Polymetallic iron oxide breccia complex: Two new polymetallic iron
oxide breccia complexes, Carrapateena and Prominent Hill, have been delineated
on the Gawler Craton in South Australia. With Olympic Dam nearby, they are pri-
marily copper deposits. Resources are in excess of 184.000 tU for Carrapateena.
Type 14—Phosphate:Chatham Rise (New Zealand), an offshore phosphorite
deposit, has been delineated and is ready to be mined. The average grade
(240 ppm U) is quite high for such a deposit.
Type 15—Black shale: Large polymetallic resources were defined in Sweden and
in Canada. In Sweden, five deposits collectively contain resources of 1300.000 tU
in the Scandinavian shales associated with V, Mo, Ni, and Zn.
2.8 Future trends
New conventional deposits: The greatest likelihood of discovering new deposits or
increasing resources is first within large known uranium provinces:
● Athabasca Basin (Canada)● Damara Belt (Namibia)● Tin Mersoi Basin (Niger)● Chu Saryssu, Syr Daria, and Kyzylkum districts (Kazakhstan, Uzbekistan)● Grants Mineral Belt and Wyoming (USA)● Elliott Lake District (Canada)● Karoo Formation (East Africa)
44 Uranium for Nuclear Power
Table 2.10 Major unconventional uranium deposits and resources discovered/defined between 2004and 2014
Country Deposit name Resources
(tU)
Grade (% U) Deposit type Deposit subtype Operator
Sweden MMS Vicken 447.755 0.0147 Black shale Stratiform Continental Precious
Minerals
Sweden Haggan 308.000 0.0133 Black shale Stratiform Aura Energy Ltd
Sweden Narke 257.000 0.0175 Black shale Stratiform URU Metals Ltd
Denmark Illimaussaq
Kvanefjeld
Sorensen
228.236 0.023 Intrusiveplutonic Peralcaline
complex
Greenland Minerals
and Energy
Australia Carrapateena 184.000 0.0155 Iron oxide breccia
● Frome Embayment (Australia)● Southern Mongolia● Chinese sedimentary basins● Thelon Basin (Canada)● Achala batholith(Argentina)● Copper Belt (RDC and Zambia)● Hoggar District (Algeria)● Central Mineral Belt (Canada)● Schist-Greywacke Group and associated two mica granites (Spain, Portugal).
Large portions of the world remain underexplored (northern Canada, Siberia,
Brazil, Argentina, Mongolia, China, central Africa, Iran, and the Arabian Platform)
containing potential targets for various types of deposits.
Most promising deposits for future planned production: Starting new mines is
essentially driven by the price of uranium and increasingly, by societal acceptance.
Thus, surficial deposits (western Australia, Namibia, Tanzania, Argentina), which are
easy to mine and sandstone deposits where uranium is recovered by low-cost in Situ
Recovery (ISR) methods (Wyoming, Mongolia, China) are the most likely to be mined
in the near future. Several intrusive anatectic deposits (low grade, large volumes, open
pits) are under development in Namibia (Husab, Valencia, Namibplass).
Unconventional deposits (co- and by-product): Very large, low-grade resources
are present around the world but very little is known about the grades and resources
of most of these deposits. Due to increasingly stringent environmental laws and
with talk of comprehensive extraction for some new projects, extraction of uranium
from very low-grade mineral deposits may bring additional uranium to the market
such as phosphorites, black shales, peralkaline complexes, etc. Today, several poly-
metallic projects are being evaluated where uranium could be extracted as a coprod-
uct: black shales from Sweden and Canada with Ni, Co, Zn, V, U, and Mo, or
deposits in alkaline intrusions from Greenland (REE, Zn, Li, U, and Th). Uranium
can also be extracted from formations such as porphyry copper deposits (Bingham
Canyon, USA; Chuquicamata, Chile) or base metal deposits (Talvivaara, Sweden).
Acknowledgments
Large portions of the data presented in this chapter come from various documents
published or being published by the IAEA. In particular, all numbers are extracted
from the uranium deposits UDEPO database. The authors gratefully acknowledge
the constructive comments of Gerard Zaluski for the improvement of the manuscript.
Books on uranium and general publications
Burns, P., Sigmond, G.E. 2013. Uranium: Cradle to Grave. Mineralogical
Association of Canada, Short Course Series 43, 440p.
46 Uranium for Nuclear Power
Committee on Uranium Mining in Virginia 2012. Uranium Mining in Virginia:
Scientific, Technical, Environmental, Human Health and Safety and Regulatory
Aspects of Uranium Mining and Processing in Virginia. http://www.nap.edu/
catalog.php?record_id513266, 345p.
Cuney, M., 2011. Uranium and thorium: The extreme diversity of the resources
of the world’s energy minerals. In Sinding-Larsen, R., Wellmer, F.-W. (Eds.), Non-
Renewable Resource Issues: Geoscientific and Societal Challenges, International
Year of Planet Earth, Springer, pp. 91�129.
Cuney, M. 2013. Uranium and thorium resources and sustainability of nuclear
energy. In: Burns, P. (Ed.), Uranium: Cradle to Grave. Mineralogical Association
of Canada, Short Course Series 43, pp. 417�438 (Chapter 15).
Cuney, M., Kyser, K., 2008. Recent and Not-So-Recent Developments in
Uranium Deposits and Implications for Exploration. Mineralogical Association of
Canada, Short Course Series 39, 257p.
Cuney, M., Kyser, K., 2015. Geology and Geochemistry of Uranium and
Thorium Deposits. Mineralogical Association of Canada, Short Course Series 46,
Dahlkamp, F.J., 2009. Uranium Deposits of the World—Asia. Berlin, Springer-
Verlag, 493p.
Dahlkamp, F.J., 2010. Uranium Deposits of the World—USA and South
America. Berlin, Springer-Verlag, 493p.
Publications in scientific reviews
Several international journals are publishing papers on ore deposit geology, geo-
chemistry, and exploration. Regularly papers on uranium ore deposits are published
in these journals. The main journals are the followings:
● Mineralium Deposita: the journal publishes since 1966 new observations, principles, and
interpretations from the field of economic geology, including nonmetallic mineral depos-
its, experimental and applied geochemistry, with emphasis on mineral deposits.● Economic Geology: the journal was first published in 1905. It publishes original papers in
any segment of economic geology and studies of all classes of mineral deposits.● Ore Geology Review: aims to familiarize all earth scientists with recent advances in a
number of interconnected disciplines related to the study of, and search for, ore deposits.
The reviews range from brief to longer contributions, but the journal preferentially pub-
lishes manuscripts that fill the niche between the commonly shorter journal articles and
the comprehensive book coverages, and thus has a special appeal to many authors and
readers.● Journal of Geochemical Exploration: is dedicated to the publication of research studies
that cover new developments in the application of analytical geochemistry and geoinfor-
matics. Themes considered by the journal include geochemical exploration and the gene-
sis of mineral deposits, environmental geochemistry and geology including metal
Uranium Mining and Milling (Issues, Industry, Impacts, Tailings), Enrichment and Fuel
Fabrication (Issues, Industry, Impacts), Depleted Uranium (Waste, Civilian and Military
Use), Tailings Dam Safety, Phosphate.● http://www.world-nuclear.org/: The World Nuclear Association is the international organi-
zation that promotes nuclear energy and supports the many companies that comprise the
global nuclear industry: uranium mining, conversion, enrichment and fuel fabrication;
reactor marketing; and nuclear engineering, construction, and waste management. WNA
provides a global forum for sharing knowledge and insight on evolving industry develop-
ments; strengthen industry operational capabilities by advancing best practice internation-
ally; speak authoritatively for the nuclear industry in key international forums that affect
the policy and public environment in which the industry operates. WNA website provides
information on nuclear energy with some 120 frequently updated information papers.● http://www.mining.com/: A source for global mining news. Since launching in February
2011, the site has 1.1 million page views per month with over 500,000 subscribers.● http://www.uxc.com/: The Ux Consulting Company (UxC) is a nuclear industry consulting
company. UxC covers the full nuclear fuel cycle with special focus on market-related
issues. UxC, founded in March 1994, publishes the Ux Weeklys and Market Outlook
reports on uranium, enrichment, conversion, and fabrication as well as publishing the
industry standard Ux Prices, referenced in many fuel contracts. UxC also prepares special
reports on key topics of interest, as well as provides data services, such as nuclear fuel
price indicator reporting, including support for the New York Mercantile Exchange
(NYMEX) uranium future contract.
References
Adams, S.S., Cramer, R.T., 1985. Data-process-criteria model for roll-type uranium deposits.
Geological Environments of Sandstone-Type Uranium Deposits. IAEA, Vienna, IAEA-
TECDOC-328.
Andersson, A., Dahlman, B., Gee, D.G., Snall, S., 1985. The Scandinavian Alum Shales:
Sveriges Geologiska Undersoekning. Serie Ca. Avhandlingar och Uppsatser IA4, NR56.
from other commodities. The decay to radiogenic isotopes is one of the aspects
exploited in the exploration for uranium deposits.
The most abundant isotopes of uranium, 235U and 238U, are commonly used in
geological sciences as geochronometers because of their long half-lives and because
they each decay to an isotope of Pb (Fig. 3.2). In deposit studies, the decay schemes
Figure 3.2 Decay schemes of 235U, 238U, and 232Th showing the position of 234U in the
decay chain of 238U, energies of the decay, and half-lives. The geochemical properties of the
decay products are quite distinct, including production of radon gas. These will tend to
separate from the ores in open systems and leak into the surrounding environment, where
they can be used to indicate proximity to a deposit.
Source: From Cuney and Kyser (2009).
54 Uranium for Nuclear Power
of 235U and 238U along with the isotopes of Pb are used to determine the age of
uranium minerals, but only 0.7204% of natural uranium is 235U. The isotope 234U,
which occurs as the decay product of 238U and is used as a geochronometer as well,
makes up only 0.0055% of natural uranium because its half-life of 246,000 years is
0.000055 times as long as the half-life of 238U. However, the decays of 235U and238U are complex, with many intermediate progeny (Fig. 3.2). These progeny decay
to other progeny along the chain until 207Pb or 206Pb are produced from 235U and238U, respectively (Fig. 3.2). Because of differences in the geochemical properties of
the progeny elements produced, they can be differentially dispersed into the environ-
ment if the system is open, resulting in disequilibrium in the radioactivity of isotopes
in the decay schemes. Some of the intermediate products such as Rn, Ra, and Bi can
be used to indicate buried mineralization, and so are useful in exploration for depos-
its. The factors that affect the deposits (so they can be detected undercover) and the
techniques used in exploration for uranium are the subjects of this chapter.
3.3 Drivers of uranium exploration
Two factors figure greatly in the exploration and exploitation of ore deposits,
namely economic factors and the role of research and technology. Although aware-
ness of economics has become more acute by most geologists because profit is
fundamental to mining companies, effective analysis of economic factors is some-
times overlooked. As a metal commodity, uranium is distinguished by its use in
energy generation, but also by its military weapon potential. The breakup of the
former Soviet Union and changes in global politics afforded the release of research
results on viable uranium deposits, such as volcanic-related uranium and metasoma-
tite, that were formerly unavailable to the West, thereby expanding the potential
importance of specific deposits that could be targeted (Cuney and Kyser, 2009).
The Organization for Economic Co-operation and Development (OECD) and the
International Atomic Energy Association (IAEA) host the Uranium Group, which
classifies uranium resources into Reasonably Assured Resources (RAR), Inferred
Resources (IR), and Speculative Resources (SR). The first two categories are often
combined into Identified Resources to reflect the realistic known resources recover-
able from uranium deposits delineated by sufficient direct measurement to conduct
prefeasibility and feasibility studies on projects. However, the true “resource” of a
commodity is a function of the resources that can be recovered within a given price
range (cf. Chapter 1). The current price is near ,USD 130/kg U (USD 50/pound
U3O8), with Identified Resources in 2013 at 5.903 million tU at ,USD 130/kg.
About 85% of RAR and IR are recoverable at ,USD 130/kg, although there is a
distinct lack of geographic diversity, with 64% of these resources located in just
five countries. Australia has the greatest resources, followed by Kazakhstan, the
Russian Federation, Canada, and Niger. However, the deposit types in each country
are different, with the iron oxide�copper�gold (IOCG) deposit at Olympic Dam
and unconformity-related deposits being the major uranium resources in Australia,
sandstone-hosted deposits as the major uranium resources in Kazakhstan and Niger,
55Exploration for uranium
volcanic-hosted and a mix of other deposit types in the Russian Federation, and
unconformity-related deposits as the exclusive resource of uranium in Canada.
Thus, exploration strategies, which are based primarily on geology and metallogeny
in the first stages of exploration, are different in each country. Most of these
resources are related to sedimentary basins, which are areas of the world where
exploration activities are most intense.
Uranium is a commodity whose value is determined by supply and demand.
Uranium supply is divided into two categories, primary supply and secondary
supply (Fig. 3.3). Primary supply includes newly mined and processed uranium,
whereas secondary supply includes highly enriched uranium from dismantling of
nuclear weapons, reprocessed uranium, mixed oxide fuels, and uranium from stock-
piles (McMurray, 2006). Primary production of uranium has been less than reactor
requirements since 1990 and secondary sources have made up the difference.
Global uranium primary production nearly met reactor requirements in 2013
(Fig. 3.3). Primary sources typically require in excess of 10 years from discovery to
production, so that after about the year 2020, the needs by commercial reactors will
have to be met by additional primary uranium supplies, new technologies for more
efficient exploitation of secondary sources, or both.
Exploration expenditures (in constant dollars) have an excellent correlation with
uranium price, with the influence of a price increase on expenditures occurring 1�2
years after the price change (Fig. 3.4).
Existing primary production Planned and existing Need
tU
90,000
80,000
70,000
60,000
50,000
40,000
30,000
20,000
10,000
0
SE
CO
ND
AR
YP
RO
DU
CT
ION
Year
1945 1955 1965 1975 1985 1995 2005 2015
Figure 3.3 Primary uranium production (in tons of uranium), production1 planned primary
sources at ,USD 130/kg and uranium needed for civilian use as a function of year,
including estimates of these until 2020. Global primary production exceeded use of uranium
until 1990. The difference between primary production and the uranium needed represents
the inventory buildup prior to 1990 or uranium derived from secondary sources since 1990
(as indicated).
Source: Data from Vance et al. (2006), Price et al. (2006), Shatalov et al. (2006), OECD
(2001, 2008), OECD/NEA-IAEA (2014), and WNA (2015a). Demand will be about 20%
greater if military and civil naval ship needs are included.
56 Uranium for Nuclear Power
Beginning in 1975, expenditures for exploration rose rapidly to a peak of USD
756 million in 1979 and then plummeted rapidly along with the spot price of
uranium (Vance et al., 2006). This rise in expenditures and price resulted first from
a combination of growth in military needs until the early 1960s, a period of time
during which the cost of uranium extraction was not controlled by economic para-
meters, and then to feed the reactors since 1973. Since 1945, the total spent on
uranium exploration has been about USD 15 billion, with the former USSR
accounting for about 20% of this and Canada only 10% (Price, 2006). Estimates of
the average historical expenditures aggregated across the exploration industry
required to find unconformity-related deposits in the Athabasca Basin of Canada
vary from USD 1.5 to 2.5 billion, or slightly more than USD 1/lb. Future discover-
ies will come at significantly higher costs as the impact of resource depletion and
current exploration technology limitations are realized.
3.4 Prospectivity, explorability, and explorationtargeting
Formulation of effective exploration strategies involves integrating (1) the prospec-
tivity of the area, (2) the economic potential of the deposit, and (3) the explorability
of the area. The prospectivity is determined by the geology and possibility of a
particular type of deposit forming, also related to the metallogeny. The economic
potential is related to the deposit type, grade, and size because this will determine
the cost of mining and hence profit margin. The economics of deposits are such
that grade is critical for deep deposits and less so for those near the surface. The
explorability is the feasibility of exploring in that area, normally related to available
300
250
200
150
Military demand
Expenditures Spot price
Energycrisis
Primary supplyexceeds demand
Supplyuncertainty
Expenditure
s c
urr
en
t $
US
(m
illio
ns)
1960 1970 1980 1990
Curr
ent $
US
/kg
Year
2000 20100
50
100
1200
1000
800
600
400
200
0
Military demand Energycrisis
Primary supplyexceeds demand
Supplyuncertainty
Figure 3.4 Spot market price of uranium in USD/kg for each year and expenditures in
millions of USD for exploration of uranium deposits as a function of year since 1960.
Source: Data from OECD (2001, 2008), OECD/NEA-IAEA (2014), and Price (2006).
57Exploration for uranium
infrastructure and politics of the host country. Countries that already mine uranium
are usually the most favorable because they have a positive uranium mining policy,
although this can change with new governments.
Most of the uranium in the world is supplied from deposits associated with
sandstone-related environments; hence, these are among those prospective for
exploration (Fig. 3.5). For example, in 2013, about 45% was by in situ leach extrac-
tion from sandstone-related deposits (ca. 12,500 tU/year), 10% was from conven-
tional mining of sandstone-related deposits, and 13% of uranium production was
from unconformity-related deposits (ca. 7000 tU/year)(OECD/NEA-IAEA, 2014).
Therefore, the most lucrative targets for uranium exploration are those associated
with sedimentary rocks, specifically those associated with Proterozoic unconformi-
ties and Phanerozoic clastic/marine basins.
In terms of mine types that have the greatest capacity potential, open pit mines
generally have greater capacity than underground mines, and underground mines
have greater capacity than mines where uranium is extracted by in situ leach technol-
ogy. However, in situ leach technology dominates current production at 51%, under-
ground mines supply about 24% of the global uranium production, and open pit
mining methods supply about 18% of the uranium produced in 2014, with the
remainder as a byproduct of other commodities (WNA, 2015b). The higher propor-
tional contribution from in situ recovery (ISR) techniques reflects a substantial
increase since 2008 and the low cost of mining using this technology. However,
deposits near the surface, even at low grades, such as the Rossing deposit in Namibia
with a grade of 0.03% uranium, can be significant economic sources of uranium.
Volcanic-related3.8%
Intrusive5.4%
Metamorphic0.2%
Granite-related1.3%
QPC4.6%
Breccia complex25.4%
Sandstone31.5%
Metasom7.4%
Surficial 3%
Collapsebreccia-type
0.1%
Phosphate2.5% Other
2.0%
Unconformity13%
Figure 3.5 Percentage of the total production (59,370 tU) of uranium as a function of deposit
type for 2013 (cf. Chapter 2).
Source: Data from OECD/NEA-IAEA (2014).
58 Uranium for Nuclear Power
Metallogeny, which is the study of the relationship between geology and the
genesis of ore deposits in space and time, is normally an important factor in formu-
lating exploration strategies. Thus, areas with favorable geology for high-priority
deposit types are desirable targets because they are prospective, although they may
not be explorable. There are uranium mines operating in 21 countries; more than
half of world production comes from just 10 mines that have average grades in
excess of 0.10% uranium, with some Canadian mines such as McArthur River and
Cigar Lake having average grades up to 25% uranium (Fig. 3.6). Countries where
uranium mining is currently active normally have both prospectivity and explorabil-
ity. Australia has the largest recoverable reserves of uranium at 29% of the global
Identified Resources at ,USD 130/kg U in 2013, in large part because of the
Olympic Dam IOCG deposit and unconformity-related deposits. Kazakhstan has
11% of the global reserves, which are mainly sandstone-type deposits mined using
0.1 kt U3 O
8
.001 kt U3 O
8
10 kt U3 O
8
1000 kt U3 O
8
Cigar Lake
McArthur River
Jabiluka 2
Unconformity related
Sandstone
Calcrete
IOCG/breccia
Metasomatic
Volcanic/intrusive
Olympic Dam
100
10
1
0.1
0.01
Gra
de (
%U
3O
8)
Ore (millions tons)
0.00001 0.0001 0.001 0.01 0.1 1 10 100 1000 10000
Figure 3.6 Grade vs tonnage for major types of uranium ore deposits. Unconformity-related
deposits, encompassed by the solid line, have the highest grade and large reserves.
Particularly high grades and reserves characterize the deposits at Cigar Lake and McArthur
River in the Athabasca Basin, and large reserves also occur in the Jabiluka 2 deposit in
Australia. Volcanic/intrusive, metasomatic, sandstone, and calcrete uranium deposits,
indicated by the group surrounded by the short dashed line, have much lower grades and are
smaller, as are IOCG/breccia related deposits, shown by group indicated by the long dashed
line, except for Olympic Dam where uranium is a byproduct.
Source: From Cuney and Kyser (2009).
59Exploration for uranium
ISR methods; Russia has 9% with a large part coming from undeveloped
K-metasomatites, sandstone, and volcanic-related deposits, and a lesser amount
from intrusive related deposits (OECD/NEA-IAEA, 2014); and Canada has the next
most plentiful at 8%, entirely from unconformity-related ore deposits.
Knowledge of the general properties of uranium, the nuclear power cycle,
deposit types, mining methods, and the economics of uranium are all important to
the uranium industry and hence to exploration strategies. To have a mine, discovery
of ore bodies is necessary, which requires a knowledge base that must be
constructed, expanded, and always refined. Most of the largest and most
profitable mines are associated with sedimentary basins, particularly those mined
using open pit methods. Finding these and other types of uranium deposits requires
knowledge of how economics in the uranium industry and research results might
affect exploration strategies, particularly as societal acceptance becomes an increas-
ing challenge for the development of new deposits all over the world.
3.5 Exploration techniques
Exploration for uranium deposits, as with any type of geological search, requires
the integration of regional geology, structural geology, geophysics, and geochemis-
try and must embrace new technologies and research results to be effective.
Although “luck” and “serendipity” will always be factors of varying proportions,
exploration must be more purposeful, especially to find deposits under cover. In
formulating exploration strategies, an analogy might be someone looking for a trea-
sure hidden in a desk drawer in a house in some unfamiliar city—geology and
metallogeny can get you in the right city and neighborhood, geophysics can get you
in the right house, and geochemistry can get you in the right room in that house.
However, you still need to find the exact drawer.
The exploration process should initially address two issues: (1) is the terrain
being considered prospective and (2) is the terrain explorable (Marlatt and Kyser,
2011)? Exploration geologists must continuously evaluate the political risk factors
as well as the viability of technical applications as projects evolve. Analysis of the
prospectivity is directed at ensuring that the exploration effort will be focused on
the best area that has the potential to host economic uranium deposits. Thus, the
metallogeny of the area—the status of the natural endowment of economic mineral
deposits in this terrain—is a significant factor. The geological assessment of the
uranium mineral endowment is a knowledge-based activity focused on understand-
ing deposit models and the probability of their occurrence.
Exploration strategies for uranium deposits vary depending on which type of
deposit is being sought. Some strategies are common to any type of deposit, such as
property evaluation involving review of the metallogeny and evaluation of available
data. Despite the diversity of deposit types, and therefore the diversity of explora-
tion strategies, some general considerations apply to exploration for all types of
uranium deposits.
60 Uranium for Nuclear Power
3.5.1 Property evaluation
Exploration means first generating targets, ranging from the right area to be in, to the
right spot on which to sample and drill. Thus, the potential property must be fully eval-
uated prior to staking or much financial investment. Many exploration geologists fail
at property evaluation not because they have not assembled all the data required for
evaluation, but because they either do not fully evaluate the data or they tend to rush
to the conclusion that the property is positive, so that the project on the property must
be promoted. Full property evaluation must be integrated and include the following:
1. Geology—Is the geology correct for the type of deposit being sought? This question
includes the tectonic environment, the timing of geologic events, and an understanding of
the metallogeny. For example, exploration for unconformity-related deposits is based firstly
on Proterozoic redbed basins overlying basement complexes and source regions character-
ized by high uranium contents. There are nearly 200 basins globally that have these attri-
butes, and about 15 of these have further attributes that make them prospective. Graphitic
metasedimentary units within the basement complex are desirable, but certainly not neces-
sary, and repeated brittle reactivations of ductile structures that offset the basal unconfor-
mity and were foci for fluid flow and ore deposition are required. The Athabasca and
Thelon basins in Canada and the McArthur Basin in Australia have favorable proven metal-
logeny for unconformity-related deposits, whereas most other Proterozoic basins are unproven
despite having similar geology. Most of them lack the metallogeny and critical data.
2. Land position availability—In areas of favorable geology and metallogeny, there must be
land that can be acquired or used with secure tenure. Some countries have formidable
land policies that can become a considerable financial burden, so knowledge of land poli-
cies must be part of the property evaluation.
3. Evaluation of historical exploration results, assessment data, and drill core data. These
evaluations should be completed prior to securing a land position. During the mini-boom
of uranium exploration in 2007 (Fig. 3.4), virtually all of the Athabasca Basin in Canada
was claimed, despite much of it being almost impossible to explore. Apparently, many in
exploration lacked the time or experience to evaluate their properties from government,
company, and published results—these must be fully evaluated and researched to avoid
re-inventing the wheel and to provide data to make intelligent exploration strategies.
4. Accessibility—Can the property be accessed by road or is a helicopter required? Assess
what types of samples (outcrop, soils, till, vegetation, drill core) can be collected as part
of the assessment.
5. Liabilities—What are the protocols for land staking and assessment requirements?
A decision to retain the property must be tempered with a complete assessment,
and only then should effort be put into exploration. Otherwise, the property should
be relinquished with minimal investment. The tendency of most companies is to
retain properties well beyond the time they should because a decision was made to
do so before the property was evaluated properly.
3.5.2 Exploration methods
Property evaluation is an ongoing and iterative process, but once all the available
data are evaluated, exploration moves into a different phase wherein data lacking
61Exploration for uranium
must be generated by the exploration team. A variety of methods is normally used
for exploration, including a balance of geological mapping, remote sensing,
geophysics, and geochemistry.
3.5.2.1 Geological mapping
Mapping is required to supplement what is available from government, industry,
and academic sources, usually at a finer scale and concentrated on the property. In
particular, mapping is focused on structures that are evident and what the structural
evolution of the area is. Structures are controlling many types of uranium deposits
related to hydrothermal fluid circulation. Structures have to be considered at differ-
ent scales. Regional scale structures or lineaments are generally transcrustal and
active over a long time period. Examples include the Arlit-In Azawa fault in Niger
for the tabular deposits of the Arlit district, the Krivoi Rog and Kirovograd shear
zones in central Ukraine for sodium metasomatic deposits, the Yuzhny fault zone
for the deposits related to K-metasomatism of the Elkon district in Russia, and the
P2 fault that hosts the McArthur River unconformity-related deposits in Canada.
Individual deposits are controlled by localized second- or third-order structures.
Structural results from mapping should be integrated with geophysical methods that
identify possible structures at depth based on physical anomalies of the rocks.
Although structure is critical in the formation of the deposits, the vast majority of
structures with seemingly appropriate characteristics are not mineralized. Mapping
is also used to characterize the cover, including Quaternary geology, soil character
and type and character of vegetation.
3.5.2.2 Remote sensing
Sensing of the physical and chemical properties of the area remotely is often part of
the property evaluation step. However, these data must be integrated into the explo-
ration strategy at a variety of scales. Remote sensing is the science of acquiring,
processing, and interpreting spectral information about the Earth’s surface and
recording interactions between matter and electromagnetic (EM) energy, mainly
from satellites. Among the remote sensing data that should be integrated in property
evaluation are digital elevation maps and light detection and ranging used to evalu-
ate the geography of the area, radiometry as part of airborne geophysics packages
or the National Oceanic and Atmospheric Administration—Advanced Very High
Resolution Radiometer and Advanced Spaceborne Thermal Emission and
Reflection Radiometer to identify radiometric anomalies within a few centimeters
of the surface, and hyperspectral data from Landsat thematic mapper that identify
types of vegetation, some mineralogy, and Fe contents (Bharti and Ramakrishnan,
2014). These techniques detect energy reflected and emitted from the Earth’s
surface from minerals, vegetation, soils, ice, water, and rocks, in selected wave-
lengths. For example, hyperspectral data have been used to characterize granite-
type uranium deposits in South China (Zhang et al., 2014).
62 Uranium for Nuclear Power
3.5.2.3 Geophysical methods
These vary with the type of uranium deposit because different aspects of the miner-
alization are targeted. The mainstay of exploration geophysics for uranium deposits
is gamma ray radiometry and spectrometry used from airborne regional to local
detailed surveys, down-hole logging, and on outcrops using handheld units (IAEA,
2013). The former is normally collected as part of airborne magnetic surveys with
magnetic contrasts used to reveal the general geology. Airborne gamma ray spec-
trometry directly measures U, K, and Th in surficial material.
Magnetics are used to map basement lithologies in basin-hosted systems.
Various EM techniques are used to image graphitic conductors that are occasionally
associated with unconformity-related deposits, paleochannels in sandstone-hosted
systems, and alteration halos in basin-related systems. For example, combined
airborne EM and magnetic surveys over the Shea Creek area on the western side of
the basin identified several conductive zones at depths of 700 m and greater that
were drilled. Ground EM surveys are used routinely to follow up airborne surveys
with deeper penetration through the sandstone, but they often provide detailed
assessments of targets of interest. Both time and frequency domain systems are
capable of deep penetration under conductive cover. Resistivity contrasts are used
to identify areas of alteration around deposits because there are differences in
porosity or mineralogy in the alteration halos. Gravity measurement reflects struc-
tures and alteration because of changes and breaks in lithology. Seismic techniques
can reveal structures, areas of alteration, stratigraphy, and depth to basement.
Emanometry has been classically used for gas detection. Radon emanometry is
based on the ability of radon (222Rn), a gaseous progeny of from 238U decay
(Fig. 3.2), to migrate to the surface from buried mineralization (Reimer, 1985).
This is facilitated through the pumping action of diurnal pressure variations and a
high permeability of the cover strata. Most instrumentation relies on alpha-particle
detection, although detection can also be done by measuring the gamma emission
from Rn decay products, 214Bi and 214Pb, following adsorption of the radon onto
activated charcoal or other substances. This technique has been used successfully
for the delineation of the Tumas calcrete-type deposit, buried under a thin cover of
calcrete or gypcrete duricrust, and the Husab alaskite-type deposit, covered by sand
dunes (Corner et al., 2010a,b), both located in Namibia. Radon emanometry has
also been used in uranium exploration for unconformity-related mineralization in
the Athabasca Basin in Canada (Earle and Drever, 1983) and the Beharchuwa-
Bokarda-Labed area of India (Banerjee et al., 2012). Helium resulting from the
alpha decay of uranium has also been used occasionally in exploration with varying
success (Clarke et al., 1983; Earle and Drever, 1983; Reimer, 1985).
Improved magnetotelluric methods have detected deep conductors and shallow
alteration zones in the search for deep unconformity-related deposits in the
Athabasca Basin (Powell et al., 2007; Tuncer et al., 2006). Clay-rich, quartz-
corroded, quartz-arenite has relatively low resistivity, whereas quartz-rich silicified
zones are characterized by high resistivity. Although expensive, three-dimensional
seismic has been used to image details of basement topology, thereby locating more
63Exploration for uranium
favorable areas for deep drilling (Hajnal et al., 2010; Wood et al., 2012) and to
detect—directly or indirectly—uranium ore using borehole seismic (Cosma et al.,
2006). In areas such as sedimentary basins where the deposits tend to be under-
cover, geophysical techniques are powerful tools to see physical anomalies under-
cover. However, it is their integration with the metallogeny and the geochemistry
that forms a sum that is greater than the parts.
Prompt fission neutron (PFN) borehole logging technology directly measures the
content of uranium in boreholes, overcoming the problem of disequilibrium, which
limits the interpretation of uranium concentrations using gamma-logging tools
(Penney, 2012). PFN uranium logging system has been recently used in the evalua-
tion of the Four Mile deposits in South Australia. The effectiveness of these meth-
ods for uranium in organic-rich terrains, such as the Canadian Shield, is
complicated by stable, soluble organo-uranium “complexes” that allow the metal to
pass through organic-rich traps.
A substantial amount of exploration dollars are normally allocated to geophys-
ical techniques in the hunt for uranium deposits to image possible mineralization or
favorable environments for deposits. Although the costs of geophysics can be sub-
stantial, drilling is much more costly than almost all of the other techniques used in
uranium exploration (Fig. 3.7).
3.5.2.4 Geochemical methods
Uranium deposits are geochemical anomalies and, as such, geochemistry must be an
integral part of the exploration repertoire. Uranium itself is commonly used as a
Dollars/km(*100)
Survey method
Dri
lling
Sels
mic
s MT
Movin
g loop E
M
IP/r
esis
tivity
Sm
all
movin
g loop
Fix
ed loop E
M
Gra
vity
MaxM
in F
-HLE
M
Rota
ry w
ing
Fix
ed w
ing
Magnetics
1000
100
10
1
Figure 3.7 Estimated costs for various geophysical surveys used in uranium exploration
relative to cost for drilling.
Source: From R.R. Koch presentation at PDAC Workshop—Giant Uranium Deposits:
Exploration Guidelines, Models and Discovery Techniques (2006).
64 Uranium for Nuclear Power
direct indicator of its deposits and this may be deflected using many different media.
Amplifying the footprint of a deposit is the purpose of using geochemistry in explora-
tion, thereby enabling detection at depth or traced near the surface (Cohen et al.,
2010; Kelley et al., 2006). There are two distinct geochemical processes that expand
the geochemical footprint of a deposit (Fig. 3.8): (1) primary dispersion, which pro-
vides information on alteration and primary element dispersion associated with ore
emplacement, and (2) secondary dispersion, which provides clues about element
migration from alteration and ore zones well after emplacement.
During primary dispersion (Fig. 3.8), components in the mineralizing fluids
permeate into the country rock, which alters primary minerals and elevates the con-
centrations of “pathfinder” elements. These enrichments relative to background
concentrations are sometimes evident in the lithogeochemistry up to several kilo-
meters away from the uranium mineralization. Hydrothermal alteration minerals
associated with mineralizing fluids are normally detected in drill core, outcrop, and
from airborne surveys of the surface using spectroscopic techniques. Enrichments
and zoning in pathfinder elements specific to an ore-forming fluid may be detected
Components from above (common Pb*, As, Cu, Zn,Ni, etc)Zone of metal
accumulation
Primary dispersion elementsinclude Ce, U, Co, Ni, S, K(and others)
Mobilized metals of distinctisotopic composition
Secondary dispersionelements include Pb*, V,As, Co, Ni, U, Some REEs,alkalis, Zn, S, Bi, Ba, He
Mineralized zones
A-horizonB-horizonC-horizionMobilized metals taken up by
clays in soils & vegetation
Anomalous Pb*, U, V, As,
REEs, Co, Bi, Ni, Ba, Zn, K, S
Water table
Figure 3.8 Diagram of buried unconformity-type uranium deposit showing elements
associated with primary dispersion (synmineralization) and secondary dispersion
(postmineralization). During secondary dispersion, elements such as Rare Earth Elements
(REEs) are mobilized from the deposit and primary dispersion alteration halo and can be
fixed by fracture fillings, clays, and Fe-Mn oxides in soils, and by vegetation. Also shown is
the influence of components deposited from above because of anthropogenic activity.
Exploration geochemistry targets both primary and secondary dispersion.
Source: After Cameron et al. (2004).
65Exploration for uranium
through element isotopic compositions that reflect the presence of ores and alter-
ation halos such as H, Li, B, C, O, S, Cu, Zn, Tl, Mo, Pb, and U or in the trace-
element compositions of specific minerals such as aluminum phosphate sulfates
(APS) and clay minerals.
The levels of uranium concentration that can be considered anomalous vary
according to the deposit type and its environment. In U-rich granites, significant
anomalies should be above some tens of ppm, the granites having themselves a
background uranium concentration of 10�30 ppm, whereas for unconformity-
related deposits, anomalies of a few ppm in the sandstone may be meaningful
because most of the sandstone has an average uranium content below 1 ppm.
In contrast to primary dispersion that is contemporaneous with the ore-forming
process, secondary dispersion occurs subsequently via the mobilization of ore or
alteration elements into the environment around the deposits. Tracing element
migration in the near-surface environment involves understanding the secondary
dispersion of elements, including those associated with anthropogenic activity that
can mask the elemental signals coming from the ores (Fig. 3.8).
Because most ore deposits are electron-rich and contain elements in reduced forms,
they are potential havens for microbes that can mobilize elements during secondary
dispersion. Microbe-mobilized elements can involve aqueous or gaseous metal com-
plexes, with the metals from the ore and the ligands from microbial waste products or
from the decay products of dead microbes. Such complexes migrate to the surface, par-
ticularly along fractures and faults, become sorbed on clay and Fe-Mn oxide surfaces in
soils, and make their way into the biosphere. These complexes have specific element
and isotope signatures that can reflect the deposit at depth (Fig. 3.8). However, the con-
trolling geochemical process by which elements migrate is fraught with uncertainty and
requires additional research. Despite these uncertainties, surface media that have been
used to trace secondary dispersion of uranium ores or alteration minerals include drain-
age sediment sampling (streams or lakes), water sampling (streams, lakes or wells), over-
burden sampling (soil or glacial tills) (McClenaghan et al., 2013; Sarala and Peuraniemi,
2007), boulder sampling and tracing (Earle et al., 1990), vegetation sampling or biogeo-
chemistry (Dunn, 2007), and rock sampling. Later mobilization of uranium from
unconformity-related deposits can be detected several hundred meters from the ore zone.
Mineral exploration in glaciated terrain has successfully used till geochemistry
and indicator mineral methods for diamonds, gold, base metals, and other commodi-
ties that reflect secondary dispersion from outcrop. There is a paucity of studies
using indicator minerals and till geochemistry for uranium exploration. A recent
study on heavy minerals in till near the Kiggavik uranium deposit in Nunavut,
Canada indicate that the highest metal contents are located directly to the west of the
deposit in locally derived, basement-dominated grey till (Robinson et al., 2014). Till
geochemistry exhibits a polymetallic dispersion signature, with anomalies in U, Bi,
Mo, Au, Ag, Co, Cs, Pb, and W up to 1 km down-ice and up-ice of the Kiggavik
Mineralized Zone. Elevated numbers of gold grains can be used as an indicator for
uranium mineralization, as can Pb-rich apatites grains. However, no uranium miner-
als were observed, perhaps because they are in a very fine fraction. Till geochemistry
surveys in glaciated areas may be a viable technique for uranium exploration.
66 Uranium for Nuclear Power
An indirect guide to uranium mineralization may be reflected by elevated
concentrations of other mobile elements such as Ba or Sr, and Pb isotopes and
iron oxidation ratio (Ng et al., 2013). Exploration success will be enhanced if
there are coincident anomalies in other elements (eg, Se, V, Mo, Pb, and Cu) that
are typically enriched in some types of uranium deposits. Many types of uranium
deposits are characterized by enrichment of one or more trace metals in addition
to uranium. Although specific tectonic environments, structural settings, and
lithologies are required for all uranium deposit types, none of these is a definitive
indicator of mineralization because in most of these settings, structures and lithol-
ogies do not host deposits. Thus, they are required for the deposits to form, but
are not definitive indicators of mineralization. In effect, the only definitive indica-
tors are physical and geochemical contrasts, both in terms of the ores and the
associated elements.
Mineralogy is an excellent exploration guide when specific alteration minerals
are developed in association with ore deposit formation, many of them can be ana-
lyzed directly in the field by portable short-wave infrared (SWIR) spectrometers
(Zhang et al., 2001) or using hyperspectral techniques (Lower et al., 2011). A typi-
cal example are the Mg-rich minerals (sudoıte, dravite), altered monazite and
zircon, and APS associated with unconformity-related ore deposits, clay alteration
around sandstone-hosted systems, and albitization associated with deposits related
to sodium metasomatism. The mineralogy of the uranium in the potential source
rocks also controls the formation of a deposit, and the mineralogy of the uranium is
also critical for evaluating the processing costs for uranium extraction.
Chemical analysis of groundwater can be a useful strategy for regional explora-
tion for uranium in reduced sediments in paleochannels, although multielement data
are required. Lead isotopes, which originate from the ore and are mobilized by
groundwaters, can be used to confirm the interpretations from the composition of
the groundwaters. Paleochannels containing uranium deposits in South Australia
have neutral, moderately saline groundwaters, whereas others are often saline and
acidic waters that preferentially mobilize radium, thereby negating the use of radon
or down-hole gamma-logging (Dickson & Giblin, 2007). If reduction of uranium by
bacteria is an effective mechanism for formation of deposits in paleochannels, then
microbially induced geochemical signals such as C and S isotopes or enhanced
mobile metals should indicate favorable areas.
Groundwater samples collected from boreholes tens of meters from
unconformity-related uranium mineralization have consistently high levels of
uranium, radium, radon, and helium (Earle & Drever, 1983). Biogeochemistry of
spruce twigs indicate that tree roots can extract anomalous uranium from ground-
water that may reflect deposits at 300 m depth (Dunn, 1984).
Lakewater and sediment geochemistry and radiometric prospecting are signifi-
cant tools in early regional exploration for uranium deposits in Canada (Cameron,
1980). This is because of the superior mobility of uranium in surface waters, which
allows the element to disperse widely from its source. Interpretation of the radio-
metric and spectrometric data must take into account that measured radiations
reflect the decay products of uranium.
67Exploration for uranium
Two sampling media in the drainage basins of lakes include organic-rich, center-
lake sediments and surface waters. Waters have certain advantages over center-lake
sediments, such as lower sampling and preparation costs. The pH of lakewater has
minimal effect on the partitioning of uranium between organic-rich sediment and water
over the pH range 5.0�7.4, but above pH 7.4 there is a marked increase in the uranium
content of lake waters relative to organic sediments (Cameron, 1980). In glaciated
terrain, such as the Canadian Shield, the development of anomalies in lakes is a two-
stage process wherein U-rich detritus is transported down-ice from the mineralized
source and then the metal is dispersed in solution from this detritus into the lakes. As a
consequence, lake anomalies are most effectively followed up by boulder tracing.
3.6 Critical factors in deposit models
Once an area has been selected because of favorable geology, metallogeny must be
considered in the evaluation process. The geology for the type of uranium deposit
must be correct, which requires knowledge of both geology and deposit models.
Associated with deposit models are certain key factors that include the timing of
the mineralization, source of the uranium, transport mechanisms of the uranium,
origin and character of the fluid, trapping mechanisms of the uranium, and controls
on paleohydrology. All of these must be understood in exploration not only to
find prospective areas, but also to eliminate areas that are unlikely to host
mineralization.
3.6.1 Timing of mineralization
The temporal evolution of the area in relation to the timing of the mineralization
must be assessed, but this is often overlooked or data are not available. Without an
idea of both the relative and absolute ages of geologic events such as faulting and
formation and alteration of the deposit and the host, exploration strategies cannot
evolve beyond being prospector-driven (Marlatt and Kyser, 2011). For example,
magmatic-related deposits form after significant differentiation of postorogenic
parent magmas to allow residual melts and magmatic fluids to become enriched in
uranium (Le Carlier de Veslud et al., 2000). In collision zones, uranium deposits
require melting of the crust, which is facilitated by crustal thickening at an appro-
priate time during convergence. Metasomatic uranium deposits are associated with
the uplift of an orogen, almost exclusively during the Proterozoic. Unconformity-
related deposits, also exclusively Proterozoic, appear to require 75�100 million
years after the basins form to allow the fluids the appropriate evolution in their
chemistry and temperature to be able to leach and mobilize uranium (Alexandre
et al., 2009). Hydrothermal processes associated with veins near granites are devel-
oped about 50 million years after the emplacement of the fertile granites. Effective
exploration in any deposit requires detailed knowledge of the time�space relation-
ships and deposit models (Cuney and Kyser, 2009).
68 Uranium for Nuclear Power
Although determining the age of uranium deposits is not straightforward, recent
advances in Inductively Coupled Plasma Mass Spectrometry (ICP-MS) and micro-
sampling technologies (Chipley et al., 2007; Kotzer and Kyser, 1993, 1995;
Ludwig, 1978, 1979; Ludwig et al., 1987), use of multiple decay systems such as
both U-Pb and Sm-Nd, precise dating of associated gangue minerals of specific
paragenesis, and integration of age data with other geologic factors render interpre-
tation of geochronologic data easier to obtain and more meaningfully for explora-
tion. The timing of the mineralizing process is required in exploration so that the
geologic, chemical, and physical environment conducive to the mineralizing pro-
cess at a critical time in the evolution of an environment can be realized.
Along with the actual age of the ores and, therefore, the critical time in Earth
history that an effective mineralizing environment was present, the timing of events
that have subsequently affected the ores can reveal when elements such as radio-
genic Pb have been mobilized from the deposits and moved into the surrounding
environment during secondary dispersion (Holk et al., 2003). These elements would
elevate element concentrations in the surrounding environment, with gradients in
concentrations and isotopes as vectors to the deposits.
3.6.2 Source, transport, and fixing of the uranium
The source of the uranium is often overlooked by many exploration geologists, but
should be a factor in evaluating areas for potential mineralization. For many depos-
its, the source of the uranium is in units that have aberrantly high uranium contents,
such as in volcanic glasses in the case of tabular deposits or in alkaline intrusions
in the case of some magmatic-type deposits. Enrichment of uranium in the source
region of basin-hosted deposits, available at the right time, certainly increases the
probability that a deposit could form. Another critical aspect is the availability of
uranium in the source region. For uranium enrichments in certain refractory miner-
als, the uranium is unlikely to be released unless those minerals become extensively
damaged by radioactive decay. In the case of sandstone-hosted deposits,
unstable volcanic glasses with high uranium contents make ideal sources because
the uranium can be effectively mobilized.
Reduced carbon, particularly in the form of organic matter, is an effective reduc-
tant for fixing uranium, given that uranium is mobile in fluids as U61 but immobile
as U41. Exactly why carbon is such an effective reductant is unclear, especially
given that there are alternatives such as sulfides and ferrous iron. Although the
latter have been shown to be the likely reductants in the formation of some depos-
its, these deposits tend not to be as large or as high grade, relative to those where
carbon is the reductant. In the case of calcrete-type deposits, redox plays a minimal
role, except as a moderator of pH. For these deposits, solubility is most important.
3.6.3 Sources of fluids
Knowledge of the nature of the fluids involved in the ore generation process and
manifestation of these fluids in appropriate environments are absolutely critical for
69Exploration for uranium
refining exploration strategies. The critical factors involved in identifying the fluids
include their temperature, pressure, oxygen fugacity, and chemical composition,
aspects that are determined using a variety of techniques during deposit studies,
including isotopic geothermometry (Kotzer and Kyser, 1995), fluid inclusions
(Richard et al., 2013), and mineral equilibria (Langmuir, 1978). Without knowledge
of the critical chemistry, temperature, and pressure required for generation of the
ores, it is difficult to identify the correct environment in which to explore. For
example, most highly evolved alkaline intrusives do not host uranium deposits
because they do not evolve to concentrate uranium in their differentiates. Most
areas of albitization do not host deposits presumably because the fluids were not
carrying uranium or there was no trap. There are areas in the Athabasca Basin with
the appropriate geology, structure, and alteration conducive to the ore-forming
process, yet they are apparently devoid of any significant unconformity-related
mineralization. Most sandstone-hosted deposits occur in meanders of paleostreams
where organic detritus could accumulate, but most such areas do not host ore.
Understanding why areas that should have ore do not, requires knowledge of the
physical, chemical, and temporal characteristics of the fluids required to form and
preserve various types of uranium deposits.
Fluids that produce many types of uranium deposits can produce significant
alteration zones around the uranium mineralization. Clay minerals are ubiquitous
up to hundreds of meters from hydrothermal uranium mineralization, and often
there is zoning in the type of alteration minerals involved. The presence of reduc-
tants such as degraded graphite or organic matter with the alteration zones are
among the major indicators of an environment conducive to uranium enrichment.
Exploration techniques that exploit these features include airborne and ground
geophysics, surface geochemistry, and clay typology.
3.6.4 Sedimentology
Most uranium deposits in sedimentary rocks are associated with geochemical
provinces enriched in U and Th or with U-rich intrusives or volcanic rocks,
although the deposits may be separated by tens of kilometers from these source
rocks. However, weak regional U and Th anomalies in sediments containing
uranium deposits may be present. Geochemical detection of uranium deposits in
sandstone-type deposits depends on the geochemical behavior of uranium and path-
finder elements (Rose and Wright, 1980). Uranium is dispersed under oxidizing
conditions, but is immobile under reducing conditions where accumulations can
occur. In addition, adsorption on Fe oxides and certain types of organic matter also
limits dispersion unless high concentrations of CO322 or other complexers are pres-
ent. Thorium accompanies uranium in most high-temperature plutonic processes,
but the two elements are separated under oxidizing conditions.
Uranium ore deposits in the Grants Mineral Belt, New Mexico occur in fluvial
sandstones in the Jurassic Morrison Formation, where uranium is concentrated by
dark grey to black humate derived from decaying vegetation. The ores vary greatly
in size and shape, generally occur in clusters, and often are difficult targets for
70 Uranium for Nuclear Power
drilling. Exploration is done primarily by drilling, delineating favorable ground on
a wide spacing, and then using closely spaced drilling in mineralized areas. Criteria
for favorable areas include the presence of a host sandstone, anomalous uranium
contents, dark color of host rock, presence of carbonaceous matter, and position of
an area with respect to mineralized trends (Fitch, 1979). Possible pathfinder
elements associated with uranium in sandstone-type deposits include S, V, Mo, Se,
As, He, Rn, and other radioactive decay products and some deposits have anomalies
in Cu, Ag, Cr, Pb, Zn, Ni, Co, Re, Be, P, Mn, and REEs. In addition, Pb, S, and C
isotopes, and textures of Fe and Ti oxides can also be vectors to ore.
In basin-related deposits, sedimentology is of outmost importance to locate the
favorable aquitards within a basin. Coarse siliciclastic continental to near-shore
sediments with detrital organic matter, confined between siltstone aquitards, are
priority targets for tabular and roll front-type uranium deposits. Specific organic-
rich sedimentary systems such as black shales or the Mulga Rock deposit in west
Australia (Fewster, 2009) exemplify the importance of basin architecture, paleo-
environment, and redox potential related to sedimentation. The importance of
sedimentology and basin architecture has also been discussed for unconformity-
related deposits (Hiatt et al., 2003; Hiatt and Kyser, 2007).
3.7 Drilling and evaluation
After targets have been identified by integrating favorable geophysical and
geochemical results, the next stage of exploration is drilling. Most alteration halos
associated with deposits are barren. Directional drilling allows several intersections
to be made from a single pilot hole, reducing drilling costs and improving target
precision. Core orientation methods are also used to better understand the structural
controls on mineralization.
Drilling results can reveal much about the third dimension of a property. Given
that drill hole samples represent an actual physical sample of the subsurface and are
among the costliest aspect of exploration, much effort should be invested to under-
stand the details of their character, mineralogy, geochemistry, and relationship to
the geological, geophysical, and geochemical results. Core or cuttings are logged to
reveal lithology and stratigraphy. Fracture densities, structure, and sedimentology
must be documented to integrate with the geological and geophysical data.
Mineralogy using SWIR or hyperspectral techniques should be an integral part, as
this will reveal the clay minerals associated with the mineralizing system relative to
background. Geochemical samples are collected, normally from the bottom of each
row of the core box, and these are composited. Although often avoided, fractures
and petrologically anomalous samples should also be collected, with the former
being compared to surface geochemical results. In addition, down-hole EM, radio-
metrics, and neutron tomography should be done if possible. At each step, the prop-
erty should continually be re-evaluated so that informed decisions about next steps
can be made.
71Exploration for uranium
3.8 Synopsis
The unique properties of uranium as a radioactive element make it both useful for
generating energy and contentious for its weapons potential. Given the existing and
projected demand for uranium, new resources must be found and new technologies
to find them and use them more efficiently must be developed.
Formulation of effective exploration strategies involves evaluating the prospec-
tivity of an area, the economic potential of the deposit, and the explorability of the
area. Countries where uranium mining is currently active tend to have both prospec-
tivity and explorability. The nuclear power cycle requires discovery of ore bodies,
which requires a knowledge base that must be constructed, expanded, and always
refined. Most of the largest and most profitable mines are associated with sedimen-
tary basins, particularly those mined using open pit methods.
Exploration strategies vary depending on which type of deposit is being sought,
property evaluation, local exploration, production rules and risks, and market condi-
tions for selection of the area to be explored. Exploration means first generating
targets, ranging from the right area to be in, to the right spot on which to sample
and drill. Thus, the potential property must be fully evaluated prior to financial
investment. Full property evaluation should include an understanding of the geol-
ogy, metallogeny, availability, accessibility, and liability. The techniques used to
evaluate a property and complete effective exploration integrate geologic mapping,
remote sensing, geophysics, and geochemistry with deposit models and the critical
aspects of the models.
New technologies will continue to be developed to help minimize the luck and
serendipity aspects of exploration. New Geographic Information System (GIS) capa-
bilities are continually expanding, as is new software to display and aid in the interpre-
tation of data. In geophysics, better resolution; refined models for EM, magnetics, and
gravity; and less expensive seismic techniques are being developed. Lower detection
limits, in situ analyses to minimize sample preparation, refined deposit models, and
isotope tracing are at the frontier in geochemistry. However, regardless of the meth-
ods, success will always depend on understanding basic geology and metallogeny, and
doing proper property evaluation.
References
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for the formation of unconformity-related uranium deposits; comparison between barren
and mineralized systems in the Athabasca Basin, Canada. Econ. Geol. 104, 413�435.
Banerjee, R., Deshpande, M.S.M., Roy, M.K., Maithani, P.B., 2012. Radon emanometry in
uranium exploration: a case study of Beharchuwa-Bokarda-Labed area, Janjgir-Champa
and Korba Districts, Chhattisgarh, India. Gondwana Geol. Mag. Spec. 13, 37�44.
Bharti, R., Ramakrishnan, D. 2014. Uraniferous calcrete mapping using hyperspectral remote
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basis of its work. The first report (ENEA, 1965) classified resources as either
Reasonably Assured or Possible Additional Resources in three production cost cate-
gories: USD 5�10, 10�15, and 15�30/lb U3O8.
4.1.2 The Red Book
Since the first publication of the aforementioned report, the ENEA and its succes-
sor, the Organisation for Economic Co-operation and Development’s (OECD’s)
Nuclear Energy Agency (NEA), has continued producing reports on uranium
resources, production, and demand, roughly every 2 years, initially with the
informal cooperation of the International Atomic Energy Association (IAEA) and
beginning in the mid-1990s with the formal participation of the IAEA’s member
states. As the majority of the editions in the series have featured a red cover, the
publication became commonly known as the “Red Book.” Over the course of its
history, the Red Book has become a recognized authoritative source of
government-sponsored information on uranium resources, and over 100 countries
have contributed data to the 25 Red Books published to date. The most recent
edition in this series, Uranium 2014: Resources, Production and Demand, was
released in September 2014 (NEA/IAEA, 2014). A compilation edition entitled
Forty Years of Uranium Resources, Production and Demand in Perspective:
“The Red Book Retrospective” (NEA, 2006) summarized information in the Red
Book series between 1965 and 2003, describing and analyzing the evolution of the
commodity and its market and drawing conclusions from its well-documented
history. Because these publications are a time series of comprehensive, global
assessments of what is known about the uranium resource base, they are the
primary references used in this chapter.
4.2 Resources and reserves
In recent years, there has been a move toward strict methods and standards for reporting
mineral resources and reserves to provide investors and governments with assurance that
amounts reported by companies are based on standard methods and are approved by
certified third party experts. Although several national and international organizations
have published codes and standards to govern reserve and resource reporting
(eg, the Australasian Code for Reporting Exploration Results; Mineral Resources and
Ore Reserves, commonly known as the JORC code; and the South African Code for
Reporting of Exploration Results, Mineral Resources and Mineral Reserves, commonly
known as the SAMREC code), they are broadly consistent. The basics are captured by
the Canadian Institute of Minerals, Metallurgy and Petroleum following the NI 43-101
code (CIM, 2014), as outlined next (italics added for emphasis).
A mineral resource is a concentration or occurrence of solid material of economic
interest in or on the Earth’s crust in such form, grade or quality and quantity that
there are reasonable prospects for eventual economic extraction. The location,
78 Uranium resources
quantity, grade or quality, continuity and other geological characteristics of a
mineral resource are known, estimated or interpreted from specific geological
evidence and knowledge, including direct sampling. A mineral resource is further
subdivided into measured, indicated and inferred categories, in order of decreasing
confidence of the estimates based on the amount of drilling and sampling
conducted.
A mineral reserve, on the other hand, is the economically mineable part of a
measured and/or indicated mineral resource. A reserve declaration includes
diluting materials and allowances for losses, which may occur when the material is
mined or extracted. It is supported by pre-feasibility and feasibility studies that
demonstrate, at the time of reporting, extraction could reasonably be justified. At
the highest level of confidence, a proven mineral reserve includes an assessment of
modifying factors, including but not limited to mining, processing, metallurgy,
infrastructure, economics, marketing, legal, environmental, social and governmental
factors. It is the kind of detailed analysis that a mining company must conduct prior
to making a decision to develop a mine to extract the well-defined reserves in the
deposit(s) of interest.
While in practice the differences are more specific and detailed, in essence a
mineral resource has the potential for economic extraction at some time, whereas a
mineral reserve is sufficiently well defined and it is understood that under the eco-
nomic conditions prevailing at the time of the declaration, the mineral(s) of interest
can be mined economically. Only resources defined to the highest level of confi-
dence (measured and indicated) have the potential to be declared a mineral reserve.
Other categories of resources typically require more sampling and analyses prior to
making a case for mine development. Because the purpose of the Red Book is to
define the amount of uranium of economic interest that could be available for use
in a global fleet of long-lived NPPs (60 years or more), uranium resource figures,
and not reserves, are reported in the Red Book.
4.2.1 Conventional and unconventional uranium resources
Uranium is relatively common in the Earth’s crust and has been recovered in a variety
of forms and settings, ranging from very low-grade to very high-grade ores that have
been mined using a number of methods. Because the focus of uranium mining has
been principally on higher-grade sources of ore where uranium is the primary mining
target, resources reported in the Red Book have been divided, somewhat arbitrarily,
into conventional and unconventional resources. Conventional resources are those
from which uranium is recoverable as a primary product, coproduct, or an important
byproduct, whereas unconventional resources refer to those from which uranium is
only recoverable as a minor byproduct, although uranium resources are huge.
Examples of conventional resources include those from which uranium minerals
are extracted and processed exclusively, from very high-grade deposits accessed by
underground mines in Canada (eg, McArthur River and Cigar Lake) to much lower-
grade ores mined by open pit in Australia (eg, Ranger) and Namibia (eg, Rossing)
and in situ leach ((ISL), sometimes referred to as in situ recovery) amenable
79Uranium for Nuclear Power
deposits in Kazakhstan (eg, Katko). Conventional resources also include multimin-
eral deposits where uranium is extracted as an important byproduct, such as the
Olympic Dam mine in Australia, where copper is the primary mining target.
Examples of unconventional resources include low-grade uranium occurrences
in phosphate rocks that are mined principally to produce phosphoric acid for
the production of fertilizer, rare earth elements (REEs), nonferrous ore, carbonatite,
black shale, lignite, and even the very low-grade background of uranium in
seawater.
The majority of the unconventional uranium resources reported in the Red Book
are associated with uranium occurrences in phosphate rocks. Although none of
these unconventional resources is currently being used for uranium supply, phos-
phate rocks and black shales have been sources of uranium in the past and interest
has not completely waned (see Section 4.4). Unconventional resources represent
significant amounts of uranium that could potentially be utilized, but the focus of
this chapter is on conventional uranium resources that will, in all likelihood, be the
source of the majority of uranium mined in the coming decades.
4.2.2 Resource classification
Uranium resources in the Red Book are classified by a two-dimensional system of
geological certainty and production costs. This classification scheme resulted from
efforts to combine resource estimates from a number of different countries into har-
monized global figures (Fig. 4.1), with amounts reported as recoverable tonnes of
uranium metal (tU). The horizontal axis of Fig. 4.1 represents the geological
Reasonably
assured
resources
Inferred resources
Identified resources Undiscovered resources
Prognosticated
resources
Reco
vera
ble
at
co
sts
Speculative
resources
<U
SD
40/k
gU
Reasonably
assured
resources
Reasonably
assured
resources
Reasonably
assured
resources
Inferred resources
Inferred resources
Inferred resources
Decreasing confidence in estimates
Decre
asin
g e
co
no
mic
att
racti
ven
ess
Prognosticated
resources
Prognosticated
resources
Prognosticated
resourcesUS
D
40
–80/k
gU
US
D
80
–130/k
gU
US
D
130
–2
60/k
gU
Figure 4.1 Red Book uranium classification scheme (NEA/IAEA, 2014).
80 Uranium resources
confidence of the estimates, from those best defined in terms of amount of drilling,
geochemical analysis, and modeling, referred to as either Reasonably Assured
Resources (RAR) or Inferred Resources (IR). Collectively, these two categories
comprise what is known as Identified Resources. Undiscovered Resources, on the
other hand, refer to resources that are expected to exist based on knowledge of pre-
viously discovered deposits and regional geological mapping. They are subdivided
into two categories: (1) Prognosticated Resources (PR) refer to those expected to
exist in known uranium provinces, generally supported by some direct evidence,
whereas (2) Speculative Resources (SR) refer to those expected to exist in geologi-
cal provinces that may host uranium deposits. Methods of calculating tonnages of
Undiscovered Resources have been available for some time (IAEA, 1992).
Undiscovered Resources provide an indication of possible resource availability
according to currently understood deposit models in areas that have been the subject
of regional geological mapping. Both PR and SR require significant exploration
efforts before their existence can be confirmed and grades and tonnages can be
defined.
Because the goal of the Red Book is to provide an assessment of global uranium
resources in a particular year, and the information collected comes from a number
of countries that, in relative isolation, developed national uranium resource classifi-
cation schemes, compromises had to be made to accommodate these well-
functioning classification schemes into a global system. This was particularly
important after the dissolution of the Soviet Union and the direct participation in
the Red Book exercise by a number of countries from the former Soviet Union. A
great deal of effort was made to incorporate regional classification systems under
the global umbrella portrayed in Fig. 4.1, while respecting the national classifica-
tion scheme definitions.
Fig. 4.2 shows the correlation of national classification schemes with the Red
Book system. Although not always a seamless fit, the Red Book classification
scheme allows major uranium resources from a variety of jurisdictions around the
world to be classified in one system.
An important aspect of Red Book resource classification is the cost of produc-
tion (vertical axis, Fig. 4.1). No other energy commodity is classified in this way.
Production cost categories were incorporated into the first edition of the Red
Book (ENEA, 1965) and are integral to the system. Governments and utilities
planning for the installation of long-lived nuclear reactors (50�60 years or more)
benefit from the assurance that potential future raw material resources will be
available at reasonable costs. Currently, RAR, IR, and SR are subdivided into
four production cost categories: ,USD 40, 40�80, 80�130, and 130�260/kgU
(roughly equivalent to ,USD 15, 15�30, 30�50, and 50�100/lb U3O8). The
highest cost category (130�260/kgU) was added to the Red Book in 2009 because
high case scenario projections of growth in nuclear generating capacity indicated
that uranium requirements could exceed identified global resources available at
costs of less than USD 130/kgU (the currency conversion of national production
costs into USD is based on the average exchange rate for the month of June in the
year of data collection).
81Uranium for Nuclear Power
When estimating the cost of production, countries are asked to take into account
the following:
● The direct costs of mining, transporting, and processing the uranium ore● The costs of associated environmental and waste management during and after mining● The costs of maintaining nonoperating production units, where applicable● In the case of ongoing projects, those capital costs that remain nonamortized● The capital cost of providing new production units, where applicable, including the cost
of financing● Indirect costs such as office overheads, taxes, and royalties, where applicable● Future exploration and development costs for further ore delineation until the ore is ready
to be mined● Sunk costs are not normally taken into consideration
All countries are asked to provide estimates of recoverable uranium resources in
the RAR and IR categories. In other words, quantities are reported as recoverable
uranium from mineable ore, as opposed to quantities of uranium contained in mine-
able ore (in situ) that do not take into account mining and milling losses. If a
4 USA 353 S. Africa 4930 Canada 3520 S. Africa 6146 USA 3420 Russ. Fed. 2760 Namibia 4503
5 Czechoslovakia 281 USSR 3800 S. Africa 3167 GDR 5245 Namibia 3221 Namibia 2715 Niger 4197
World
Total
4804 46,290 37,736 69,692 50,026 35,755 54,670
% by
top 5
94% 81% 83% 73% 67% 75% 77%
Notes: DR Congo, Democratic Republic of the Congo; GDR, German Democratic Republic; USSR, Union of Soviet Socialist Republics; USA, United States of America; S. Africa, Republic of South Africa;
Sources: for 1950�2000, OECD-NEA (2006); for 2010, OECD-NEA/IAEA (2012).
mine apply. In general, the higher the grade and the larger the deposit, open pit
mining is economically employed to greater depths than would be considered for
lower-grade or smaller deposits. Hence, a large open pit for uranium may be over a
kilometer across and hundreds of meters deep, for example,
● Rossing mine, Namibia, mined 1976�present: 3000 m long, 1500 m wide, and up to
390 m deep (Rossing Uranium, 2014, see Fig. 6.1)● Lichtenberg open pit in Saxony, Germany, mined 1956�1976, now backfilled: 2000 m
long, 1000 m wide, and up to 240 m deep (Wismut, 2014)● Ranger Pit 3 in Australia, mined 1997�2012 and currently being backfilled: 1150 m long,
800 m wide, and up to 275 m deep (Wines et al., 2013)
Open pits must also be protected from flooding from nearby catchments or
streams, and may also require control of groundwater inflows.
Over the history of particular deposits, open pit mines have been developed
where underground mines formerly existed, and open pit mining may be followed
by underground mining to access deeper ore.
6.2.1 Soft rock open cut mining
Rock properties are a major determinant of mining method. Where the hosting rocks
are relatively friable (low geotechnical strength), rock material may be removed by
scraper technology, where the equipment excavates, transports, and dumps material.
Figure 6.1 Rossing open pit uranium mine, Namibia, April 2012.
Source: Photograph P. Woods.
128 Uranium for Nuclear Power
Generally, only a thin layer of material is removed with each pass. Alternatively,
material can be loosened and moved by the blade of a bulldozer and pushed into a
heap. The heap is then either loaded onto a truck or conveyor belt for removal, or in
some cases mixed with water and made into a slurry that is pumped away.
For stronger material, ripping by large tines mounted on a grader or bulldozer
may be used to break up the material before it is removed.
Alternatively, “free digging” material can be dug with an excavator without any
additional assistance. The excavated material is placed in a truck or conveyor belt
for removal. Selective mining of ore is proposed to be by a “surface miner” that
removes and breaks up a relatively shallow strip of ore for direct loading onto a
truck, and produces a “precrushed” product for further processing (Toro Energy
Limited, 2015).
6.2.2 Hard rock open cut mining
Material with higher strength must be broken up or fragmented prior to excavation
and removal. The most common method is drill-and-blast. A drill rig, or group of
rigs, prepares holes, usually vertical, at a predetermined spacing and pattern, which
are then loaded with explosives, usually a powder, pellets, or a packaged form. The
explosives are detonated in a defined sequence designed to fragment the rock ready
for excavation and placement onto a truck or conveyor for removal from the open
pit (Fig. 6.2).
6.2.3 Trends and alternatives
Advancements in mining machinery have meant deeper and deeper open pits have
been viable over the last century. Small open pits dug by hand are known histori-
cally. The optimum size of mining, rock, and ore hauling equipment is largely
Figure 6.2 Drill-and-blast (left; Langer Heinrich mine, Namibia) and excavation
(right; Rossing mine, Namibia) at hard rock open cut mines.
Source: Photographs: (left) P. Waggitt, reproduced with permission, (right) A. Hanly,
reproduced with permission.
129Uranium mining (open cut and underground) and milling
determined by the required rates of mining, the length of mine life, and the dis-
tances waste rock and ore must be hauled. These trends are apparent in uranium
mines as much as any other industrial-scale mines and, with the exception of addi-
tional dust control to reduce radiation exposure where required, general mining
industry planning and machinery are applied at open pit uranium mines.
Old tailings can sometimes be recovered by hydraulic mining or sluicing, where
high-pressure water jets are used to both break up the tailings into slurry that can
then be pumped for further processing (see Section 6.5.1).
6.3 Underground mining
Underground mining is employed to access ore that occurs below the ground sur-
face. Hand-dug underground mines for minerals other than uranium are known
since antiquity. Underground uranium mines, or mines with a significant uranium
byproduct, have all been undertaken in the era of mechanized mining.
6.3.1 Access options
Access to an underground uranium orebody is by one or more means, and the
design depends on the topography of the ground surface, the depth of the orebody,
rock strength conditions, and possible requirements for road access. The three main
access methods are shafts, adits, and declines.
Shafts are vertical openings typically a few meters in diameter or width. Shafts
may be tens or hundreds of meters deep, or over a thousand meters in some cases.
Shafts may start at the surface or be created between different levels within an
underground mine. Shafts are typically lined with concrete (or historically with tim-
ber); although in strong, competent rock they may be unlined. A combination of
lined and unlined shafts is possible in changing rock strength conditions. Lining or
grouting may also be required to control the inflow of groundwater, even if rock
strength would otherwise allow an unlined shaft. Materials and miners are trans-
ported in cages or large containers supported on cables operated by a winch beneath
a headframe (Fig. 6.3).
Adits are near horizontal openings, typically from the side of a hill or mountain
to access an orebody at a similar elevation. Again, they are typically a few meters
in the vertical and horizontal dimensions, although in early mines the openings
might only be 1.5�2 m in size. A railway may be installed in an audit to transport
material and miners, or access may require using wheeled vehicles. Conveyors sys-
tems might also be installed in an adit for the removal of ore and waste rock. As
with shafts, lining with concrete, steel, or timber may be required if potential col-
lapse is a risk at the particular site. An adit may also act as a drain for groundwater
that enters a mine, in which case it would slope gently upward from the entrance.
Declines are inclined ramps, often in the form of a spiral, constructed to allow
vehicular access to the orebody. The entrance to a decline is typically called a
130 Uranium for Nuclear Power
portal. Declines are typically no steeper than a 15% slope. Most of a decline is
wide enough only for the largest vehicle that uses it, so passing spaces must be pro-
vided at various intervals to allow two-way traffic. Again, lining may be required,
depending on rock strength conditions.
These access options are often used concurrently: typically, a mine can have
multiple shafts with one or more adits or declines, depending on the size and situa-
tion of the underground workings that require access. Where a single access point
is used, particular attention to emergency access is required, should that access
become blocked, such as via a ventilation shaft (see Section 6.3.3).
6.3.2 Mining options
Individual mining options are diverse, and new techniques are developed as tech-
nology advances or specifically to meet particular ore geometry or characteristics
not economically amenable to established techniques. Four commonly used techni-
ques are described here; the interested reader is referred to the extensive mining lit-
erature or to descriptions of individual mines for more information.
Further to the access from the surface, mining areas are accessed by horizontal
or inclined tunnels and internal shafts, and small railways may be constructed or
material moved in specialized underground trucks. In large, long-lived mines, the
system of tunnels may be hundreds of kilometers long in total. The mining first
described here presumes competent rock material; for an example of specialized
mining in weaker material; refer to the end of this Section.
Where ore occurs in relatively narrow veins, small scale, labor-intensive techni-
ques may be required. Usually drilling and blasting is required to break up the ore
for excavation and removal. This type of mining may proceed from the base of the
orebody upward, following the slope of the ore, from above proceeding downward,
or a combination. Pillars of unmined ore may be required to hold the mining
void open, or full removal may be possible. The void left behind may remain open,
Figure 6.3 Left: Headframe at Dolni Rozinka uranium mine, Czech Republic, with some
timber supports awaiting transport underground; right: headframes at shafts 1 and 2, Cigar
Lake uranium mine, Canada.
Source: Photographs: left, P. Woods; right, Cameco; reproduced with permission.
131Uranium mining (open cut and underground) and milling
be allowed to collapse, or be partly or completely refilled with waste rock or tail-
ings from an associated milling facility.
A current example of a small underground uranium mine using narrow vein min-
ing techniques is Dolni Rozinka in the Czech Republic, which extends to more than
1000 m below the ground surface (Michalek, 2011) and there are many historic
examples around the world. In larger mines, several areas of mining may be active
at once to provide sufficient quantities of ore for further processing.
If the ore mass is larger, a typical technique used is called stoping. There are a
number of variations, but it is used where the orebody can be mined in a series of
vertical masses that might be tens or more meters high and several to tens of meters
wide. Access tunnels are made and the selected portion of orebody is drilled and
explosives placed. After blasting, the broken ore is removed from the base of the
blasted area for crushing and processing. Previously mined stopes may be refilled
with waste rock or tailings, often with the addition of cement. Once the refilled
stope has gained sufficient strength, the adjacent segment or segments of the ore-
body can be mined. Again, multiple stopes may be mined simultaneously, or be in
different stages simultaneously, to provide the required rate of ore for further
processing.
A current example of a large underground mine using stoping is the Olympic
Dam copper�uranium�gold mine in South Australia (Robertson et al., 2013;
Webb, 1998; Oddie and Pascoe, 2005).
Where the orebody is relatively horizontal and extensive, ore is typically
removed in galleries, leaving sufficient columns to prevent collapse—this is called
“room and pillar” mining. When an area has been mined to the fullest extent possi-
ble using this method, the space can either be backfilled to allow mining of the pil-
lars, or the pillars can be progressively removed to allow collapse of the mined-out
area. An example of an underground uranium mine using this method with backfill-
ing is COMINAK’s d’Akouta mine, Niger (Bibert, 1982; Areva, 2015).
Some of the high-grade uranium mines in Canada operate in relatively unconsol-
idated sediments. The Athabasca Basin is at a high latitude and ground freezing can
be used to isolate the ore from the surrounding highly hydraulically conductive, sat-
urated sediments that would otherwise cause large-volume groundwater inflows and
possibly collapsing conditions (Newman et al., 2011). Ore is extracted by the “raise
bore” technique, such as with the McArthur River mine (Beattie and Davis, 2002;
Jamieson, 2002). Tunnels are first constructed above and below the orebody to be
mined. A relatively narrow vertical hole (bore) is drilled between the two levels.
The raise borer, essentially a cutting disk, is rotated raised upward starting at the
lower level, increasing the diameter of the bore to (for example) 3 m. The extracted
ore is removed from below by a remote-controlled bucket loader and delivered for
scanning of ore grade and either further milled underground and pumped as a slurry
to the surface, or hoisted above ground for further milling (Beattie and Davis,
2002). At the newer Cigar Lake uranium mine, a variation of this technique is being
used called “jet boring.” Here, the orebody is artificially frozen, and rather than a
mechanical device, rotating high-pressure jets of water are used to both loosen the
ore and to create a slurry that can be collected and pumped away for processing
132 Uranium for Nuclear Power
(Schmitke, 2004; Edwards, 2004; Fig. 6.4). Only a tunnel beneath the ore is
required. Mined voids are filled with a concrete mixture to allow later mining of
adjacent ore.
6.3.3 Ventilation considerations
Ventilation is always important for underground mines, and this requirement must
be a special consideration for uranium mines, or mines with significant uranium
associated with them. This has been recognized from the early years of uranium
mining (Westfield et al., 1958), although those authors acknowledged that at that
Freeze plant
Surface freeze pipes
Underground
freeze pipes
Brine
circulation
Sandstone
Unconformity
Shuttle car Rod cars Drilling
car
Slurry
car
Clamshell
bucket
To
underground
processing
ROM
Drawing not to scaleBasement rock
Uranium orebody
High-pressure
water jet
Mining cavity Backfilled cavities Cameco
480-m levelHeat
exchangers
Temperature holesfrom surface to
underground sensors
at various elevations
Figure 6.4 Ground freezing and jet boring mining technique, Cigar Lake, Canada.
Source: Courtesy Cameco, reproduced with permission.
133Uranium mining (open cut and underground) and milling
time, “Too many underground uranium mines are operating with inadequate natural
ventilation” (p. 10).
Uranium itself is radioactive, but most of the radioactivity associated with ura-
nium ore is from the radioactive decay products. Of these, radon gas and its decay
products (which may remain airborne for some time) and radionuclides associated
with dust must be taken into account. A modern account of ventilation requirements
is presented by Gherghel and De Souza (2008). Ventilation is forced by fans and
controlled within a mine by special ventilation shafts and internal mine connections,
as well as through general access shafts and tunnels. Special doors and vents can be
opened or closed to direct air where it is needed. Personal protective equipment
such as a respirator of dust mask may still be required for some tasks such as dril-
ling. Computer modeling is required, undertaken together with in-mine measure-
ments (eg, Bracke et al., 2006; El-Fawal, 2011).
As an example, at the high-grade McArthur River operation in Canada, the
primary ventilation system is driven by surface exhaust fans on shaft no. 2 that
draw fresh air (heated in winter) through two other shafts (Cameco, 2009). The ven-
tilation is required to control diesel emissions and blasting gases, as at nearly all
underground mines, as well as radon and long-lived radioactive dust (LLRD).
General ventilation principles followed at McArthur River consist of the following
(Cameco, 2009):
● Single pass ventilation used in areas with a significant radon or LLRD potential● Capture and containment of radon at the source location (drill collars, radon bearing
sumps, backfill holes, etc.) through the use of secondary ducting● Equipment operator stations will be located in fresh air and upstream of potential radiation
sources whenever possible● Avoiding the short-circuiting of air from fresh air to exhaust air
Other documented recent case studies include the Indian Jaduguda uranium mine
(Panigrahi et al., 2005) and at Olympic Dam in Australia (Uggalla, 1999; Bloss,
2009).
Occupational radiation protection in the mining and processing of raw materials
is the subject of one of the International Atomic Energy Association (IAEA)’s main
safety guide series documents (IAEA, 2004). Even with relatively low ore grades,
along with silica dust and diesel fumes, radiation protection requirements may be
one of the main drivers of ventilation at an underground uranium mine, and similar
efforts may be required in mines not currently producing uranium as a byproduct,
but with significant uranium present in the orebody.
6.3.4 Identification and placement of wasteand mineralized material
The uranium grade of material in a pit is generally known in advance from a con-
ceptual or computerized model of the orebody and waste rock, and by testing of
some or all blast holes (where used) by analysis of samples, geophysical probing
(using natural gamma), or both or by surface gamma scanning. Calibration of
134 Uranium for Nuclear Power
gamma loggers is important and can be difficult, particularly if grades are low and
disequilibrium between uranium and the decay products responsible for the gamma
is present (eg, Princep and Owers, 2012).
It is common for broken rock to be passed under a discriminator, that is, one or
a set of gamma ray detectors that have been calibrated for local conditions. The dis-
criminators provide an estimate of the grade of ore in that truck, so the truck can be
directed to dump its load directly into a crusher or to the relevant stockpile of ore
grade, mineralized material but below current ore grade, or clean waste rock
(Fig. 6.5, cf. Tietzel et al., 2013 at Ranger Mine, Australia).
Rock may also need to be specifically placed to control dust generation, the genera-
tion of acidic or otherwise potentially toxic water runoff, or seepage to groundwater.
Other criteria are required to separate potentially problematic rock, and a carefully con-
sidered plan to manage it must be prepared and implemented (eg, INAP, 2014).
Historically, with poor quality water from early uranium mines designed without ade-
quate environmental safeguards, the major problematic contaminants have not always
been radioactive; rather, other elements and ions such as heavy metals (eg, by acid
rock drainage) or salinity have more often been of most concern1.
6.4 Milling and extraction
The milling and extraction of uranium from uranium ore is a mixed physical
and chemical process. A generalized flow chart of the major steps is outlines in
Fig. 6.6. A compendium of uranium extraction technology was published by the
Figure 6.5 Discriminator in action at Ranger uranium mine, Australia.
Source: Dr. J. Kvasnicka, Radiation Dosimetry Systems, Adelaide, Australia, reproduced
with permission.
1Committee on Uranium Mining in Virginia; Committee on Earth Resources; National Research
Council (2011)
135Uranium mining (open cut and underground) and milling
IAEA (1993) and earlier overviews by Seidel (1980) and the IAEA (1971). The
more recent reviews by Edwards and Oliver (2000), Gupta and Singh (2003, Ch. 3
& 4), Lunt et al. (2007), Taylor (2011), Mishra et al. (2013), and Schnell (2014)
provide useful summaries of uranium processing; Mishra et al. (2013) and Edwards
and Oliver (2000) give more depth, whereas the others are useful but briefer practi-
tioners’ overviews.
6.4.1 Conventional crushing and grinding (comminution)
In most instances, uranium ore as produced by a mine—run-of-mine (ROM) ore—
contains too much coarse material for further processing by comminution. Crushing
(producing fragments typically of a few centimeters size) and grinding (producing
sand or finer sized particles) are generally employed. Both are mechanical pro-
cesses; crushing is generally a dry process, whereas grinding is typically a wet pro-
cess. Because of the dust generated by crushing, dust control by water sprays or
other means may be required for occupational health and visibility reasons.
Open pit
mining
Crushing &
grinding
Leaching
Separate solidsTailings disposal
In situ
leach mining
Extract U
in liquor
Precipitate
uranuim
Recycle
barren liquor
Separate solids
Drying
Uranium oxide concentrate, U3O
8 (yellowcake)
contains approximately 85% by weight of uranium
Recycle
barren liquor
Underground
mining
Figure 6.6 Generalized flow sheet for the processing of uranium ore. Not all variants of
processing are shown.
Source: World Nuclear Association, reproduced with permission.
136 Uranium for Nuclear Power
Note that “mill” and “milling” may be used to refer to just the comminution pro-
cess of a mineral processing plant or to the plant as a whole, including comminu-
tion (or even in the absence of comminution), extraction, and purification.
The sizes of crushing and grinding plants must be matched to the type of ore
(hardness, size of ROM ore, clay content, etc.) and the throughput of the proces-
sing plant. The purpose of crushing and grinding is to liberate or at least expose
the surfaces of uranium minerals to allow the next stage of extraction. The num-
ber and design of mills and the optimum fineness of the grind will vary accord-
ing many mineralogical, physical, metallurgical, and commercial factors. Finer
grinding is typically required for alkaline leaching projects compared to acid
leaching (see Section 6.4.3). Grinding is undertaken in large, rotating drums and
may be enacted by the tumbling of rock pieces themselves (autogenous grinding)
or with the addition of steel rods, steel balls, or rock pebbles to facilitate the
grinding process (see Fig. 6.7).
Other factors can affect the design of a crushing and grinding circuit. For some
clay ores (and other reasons), crushing may not be effective and the broken ore is
Figure 6.7 Rod mill with steel grinding rods ready to load, Namibia.
Source: Photo: Rossing Uranium Mine, used with permission.
137Uranium mining (open cut and underground) and milling
sent immediately to a grinding process (eg, Rabbit Lake in Canada in the 1980s;
Woods, P. 2015. Environmental and social aspects of feasibility studies, mining operations
and closure; balancing realities and expectations from different angles. In: Proceedings
ALTA 2015, Perth, Western Australia, 23�30 May 2015.
Woods, P., Edge, R., Fairclough, M., Fan, Z., Hanly, A., Miko Dit Angoula, I., et al. 2015.
IAEA initiatives supporting good practice in uranium mining worldwide. In: Merkel, B.,
Arab, A., (Eds.), Uranium—Past and Future Challenges (Proc. Uranium Mining and
Hydrogeology VII, Freiberg, Saxony, Germany, 21�25 September 2014), pp. 31�39.
156 Uranium for Nuclear Power
7Introduction to uranium in situ
recovery technology
Mark S. Pelizzaa,1 and Craig S. Bartelsb,2
aM.S. Pelizza & Associates LLC, Plano, TX, United States, bHeathgate Resources Pty Ltd,
Adelaide, SA, Australia
7.1 General description
7.1.1 The ISR technique
In situ is a Latin word that translates literally to “on site” or “in position.” Unlike
conventional mining methods, where the uranium mineral and host rock are exca-
vated together and the uranium is recovered on the surface, in situ recovery (ISR)
technique removes the uranium while leaving the host rock in place. ISR utilizes
wells to inject amended groundwater into the ore zone. This groundwater solution
is commonly referred to as lixiviant. The lixiviant dissolves uranium as it is drawn
through the uranium-bearing host rock by a pump in a nearby production well,
which then sends the uranium-rich water to the processing plant where the uranium
is recovered. The water is then refortified and sent back to the ore zone through the
injection wells to recover more uranium. The cycle continues until the desired ura-
nium extraction is complete.
ISR3 involves the entire recovery process, in which a fortified groundwater leaching
solution (lixiviant) is used to extract the uranium from the geologic formation in which
it occurs and then the recovered uranium is concentrated and packaged into a form that
1Mr Pelizza’s career in the uranium industry has spanned 36 years, with 34 years in various positions
with Uranium Resources, Inc. where he most recently served as Sr. Vice President of Environment,
Health, Safety and Public Affairs. Currently, Mr Pelizza is a consultant to the uranium industry through
his firm M.S. Pelizza & Associates LLC. He serves as a Director with enCore Energy Corp. Mr Pelizza
holds a BS degree from Fort Lewis College and a MS from Colorado Sc. of Mines. He is a registered
Professional Geoscientist in Texas. He has served as President of the Uranium Producers of America,
and Chairman of the Texas Mining and Reclamation Association.2Mr Bartels has worked in ISR (uranium and copper) for 37 years with various companies, and has been
involved in all aspects of ISR during that time (well field design and operation, resource estimation,
process facilities design and operation, company management). He is currently President of Heathgate
Resources Pty Ltd in Australia, which owns and operates the Beverley uranium ISR operation, and is
operator of the Four Mile ISR project.3 In situ recovery (ISR) is commonly called in situ leaching (ISL). ISR is the commonly used term in the
United States. In this chapter, the term ISR is appropriately used because the entire uranium recovery
cycle is described from subsurface dissolution, through processing, drying, and packaging.
Uranium for Nuclear Power. DOI: http://dx.doi.org/10.1016/B978-0-08-100307-7.00007-7
can be sold for nuclear fuel. This is accomplished by injecting the lixiviant through
injection wells completed in the zone of interest, leaching the target minerals, recover-
ing the uranium-rich lixiviant by pumping production wells and then processing by ion
exchange (IX), chemical treatment, drying, and packaging (Fig. 7.1, Photo 7.1).
Laboratory
Controlroom
Powerstation
Evaporationponds
Trunk lines
Willawonina
formation
Shoestring sands,
clays and gravels
Beverley clay
Beverley sands Uranium
Alpha mudstone
Well house
Productionmonitor well
Uranium extraction columns
Uranium recovery columns
Thickeners Yellowcake dryingand packaging
Shippingcontainers
WorkshopExtractionfiltersReagentstorage
Overlyingmonitor well
Productionmonitor well
Injection well
Extraction well
Monitor well
Screened interval
Injected solution
Uranium enriched solution
Figure 7.1 Conceptual block diagram of the ISR technique.
Source: World Nuclear Association/Heathgate Resources.
Photo 7.1 ISR well field at the Beverly ISR project, Australia.
Source: c.2004, Heathgate Resources Pty. Ltd.
158 Uranium for Nuclear Power
Various well patterns are typically used for uranium ISR. Each well field area
consists of groups of these patterns that are installed to correspond to the specific
geometry of the orebodies.
To be amenable to ISR, the lixiviant must be able to contact the uranium mineral in
the host rock, the uranium mineral must be a species that is soluble to the specific lixi-
viant, and there must be adequate groundwater and transmissivity to allow for hydro-
dynamic control. Generally, this means that uranium ISR is conducted in saturated
permeable sandstone aquifer systems that are rich in redistributed uranium minerals.
The natural groundwater in these aquifer systems is the fundamental component of the
lixiviant and varies from to low total dissolved solid (TDS) “fresh” water to high TDS
“saline” water that would be unsuitable for consumption (Hunter, 2001). However, it
is important to note that even in freshwater aquifers, in the mineralized portion of the
aquifer system that will be subjected to ISR, the groundwater is not potable because
the concentrations of uranium and uranium progeny such as 226Ra and 222Rn exceed
acceptable drinking water standards by a large margin (Pelizza, 2014).
Depending on regulatory requirements and natural geotechnical variables, an
acidic or an alkaline lixiviant (leaching solution) may be used at a given location.
Typical acidic lixiviants consist of native groundwater to which sulfuric acid and per-
haps an oxidant is added. A typical alkaline lixiviant consists of native groundwater
to which gaseous carbon dioxide (or some form of sodium bicarbonate) and oxygen
(or other form of oxidant) are added. After injection of the barren lixiviant into injec-
tion wells, and recovery through production wells, it is piped to the IX facility, where
the uranium is removed by circulating the uranium-rich lixiviant through IX resin.
The barren lixiviant will then be returned to the well field. When loaded with uranium
to capacity, the IX is washed of uranium and the uranium is further processed to its
final form. IX resin is returned to service after it has been stripped of uranium.
The chief components of an ISR project include the leaching process and the
plant in which the uranium is processed. In the leaching process, a lixiviant stream
is continuously recirculated from the recovery plant into injection wells through the
ore bearing strata. From this, a uranium-rich lixiviant is withdrawn (via production
wells), pumped through the recovery plant [IX or, less often, solvent extraction
(SX)],4 refortified with reagents, and circulated to the injection wells. In the proces-
sing plant, the uranium is removed (eluted) from the IX resin, or SX circuit, and
further treated (precipitated, dried, packaged). The IX process may be integral to
the entire recovery process or remote at satellite plants and the resin transported to
the remaining recovery steps. Acid consumption depends greatly on minerals other
than uranium in the orebody.
7.1.2 Advantages of ISR
Uranium ISR is a proven technology that has been successfully demonstrated com-
mercially worldwide. ISR of uranium has environmental advantages over conven-
tional open pit uranium mining, as evidenced by the following.
4 Ion exchange (IX), rather than solvent extraction (SX), is more generally considered here since it is, by
far, the most prevalent method in ISR for removing uranium from lixiviant.
159Introduction to uranium in situ recovery technology
● ISR results in significantly less surface disturbance. Mine pits, waste dumps, haul roads,
and tailings ponds are not needed.● Compared to conventional mining, ISR reduces the short- and long-term exposure of the
general population to the extremely low levels of radioactivity because almost all of
the source term (notably the radioactive decay products of 238U) remains underground in
its natural location. Very little residual radioactive waste is produced and there are no tail-
ings. Land and water are returned to their original pre-ISR uses and quality.● ISR requires much less water than pit or underground mine dewatering, or conventional
milling.● Minimal use of heavy equipment, combined with the lack of haul roads, waste dumps,
etc., results in virtually no air quality degradation at ISR sites.● Following the initial construction activities, fewer employees are needed at ISR sites,
thereby reducing transportation and socioeconomic concerns.● Aquifers are not excavated, but remain intact during and after ISR so after any required
rehabilitation they remain available for future uses. Avoiding the creation of large excava-
tions preserves the surrounding land for grazing, raising crops, and other traditional uses.● The technology of recirculating groundwater through the IX facility reduces the amount
of solids to a negligible quantity, and tailings ponds are not used, thereby eliminating a
major groundwater pollution concern.
7.1.3 Overview of global ISR operation
Uranium produced by ISR accounts for 46% of the uranium produced worldwide
(World Nuclear Association, 2015a). Major historic and current production centers
and operations are found in Australia, Asia, Eastern Europe, and the United States,
as summarized next.
7.1.3.1 Australia
The Beverley in South Australia (520 km north of Adelaide, on the plains northwest
of Lake Frome) was Australia’s first ISR project, starting operation late in 2000.
The project uses sulfuric acid leach chemistry. It was licensed to produce 1000 tU/
year (2.24 mm lb U) and reached this level in 2004, though production has declined
since. Mining of Beverley ceased at the end of 2013 (World Nuclear Association,
2015b).
The Four Mile leases are contiguous with Beverley, and ISR of the east orebody
commenced in April 2014. Uranium recovery is through the neighboring Beverley
Pannikan satellite IX plant with the loaded resin transported by truck and trailer
to the main Beverley plant for stripping (elution) and precipitation, as is done at
many US projects. Production is currently at about 1000 tU/year (2.24 mm lb U)
(Fig. 7.2, Photo 7.2).
The Honeymoon ISR project in South Australia commenced operation in 2011,
and, because of the very high sulfate levels in the native groundwater, has been one
of the few ISR projects to use SX, rather than IX, to remove uranium from well
field lixiviant. In November 2013, the project was closed and put on care-and-
maintenance until uranium prices improved.
160 Uranium for Nuclear Power
Photo 7.2 The Beverly ISR process facility (2001), Australia.
Source: Heathgate Resources Pty. Ltd.
Figure 7.2 Australia uranium mines.
Source: OECD (2014).
161Introduction to uranium in situ recovery technology
7.1.3.2 Asia
China. Prior to the 1990s, China’s uranium resource development activities were
mainly carried out on hydrothermal-related granite-type and volcanic-type uranium
deposits in the Jiangxi, Hunan, and Guangdong provinces and the Guangxi
Autonomous Region of Southern China. Facing the challenge of meeting demand
for economic uranium resources for China’s mid-term and long-term nuclear energy
development plan, the direction was changed from “hard rock”-type deposits to ISR
amenable deposits in Northern and Northwest China (OECD, 2014). ISR began in
1993 at the Yining production center Kujieertai deposit using sulfuric acid chemis-
try. Reported production in 2012 was 380 tU (.84 mm lb U) (OECD, 2014). The
Shihongtan deposit, located in Tuha basin of Xinjiang, is a new uranium discovery.
Preliminary ISR field testing began in 2000, including lab tests and field tests.
During field testing with sulfuric acid, numerous problems were encountered
because of the high content of CO2 in the ore. Therefore, a NH4HCO3 ISR test was
to be conducted (Xuebin et al., 2004) (Fig. 7.3).
Kazakhstan. Uranium exploration began in the 1940s in Kazakhstan, making it
one of the first countries to develop uranium resources (Kim, 1997). In 2009,
Kazakhstan became the world’s leading source of mined uranium, producing almost
28% then, 33% in 2010, 36% in 2011, 36.5% in 2012, and 38% in 2013, almost all
using ISR methods in recent decades. ISR uranium production in Kazakhstan
Figure 7.3 China uranium mines.
Source: OECD (2014).
162 Uranium for Nuclear Power
requires large quantities of sulfuric acid due to relatively high levels of carbonate in
the orebodies (World Nuclear Association, 2015c) (Table 7.1).
Mongolia. From 1945�60, numerous uranium occurrences were discovered in
Eastern Mongolia. Currently no uranium is being produced in Mongolia; however,
a number of projects are in the planning stage of development, with start dates at
the end of this decade. The following deposits are planned for development using
sulfuric acid chemistry: the Kharaat, Khairkhan, Gurvansaikhan, and Ulziit deposits
at the Gurvansaikhan Production Center; and the Dulaan Uul and Zoovch Ovoo
deposits at the Coge-Gobi Production Center (OECD, 2014) (Fig. 7.4, Photo 7.3).
Russia. According Fazlullin et al. (2004), uranium ISR is performed in Russia on
two deposits, the Dalmatovskoe/Khohlovskoe and Khiagdinskoye. The Dalmatovskoe/
Klolovskoe deposit is located in the Trans-Ural region. The annual uranium
production in 2012 was 578 tU (1.28 mm lbs) (World Nuclear Association, 2015d).
The Khiagdinskoye deposit, the only ISR deposit in permafrost, is located within
the Republic of Buryatia (World Nuclear Association, 2015e). Multiwell experi-
ments began there in 1999 and in 2002 50 tU (110 m lbs) were produced at about
Table 7.1 Kazakh uranium production by mines (tU) (WorldNuclear Association, 2015a)
Province/
Group
Mine 2010 2012 2013
Chu-Sarysu,
Eastern
Tortkuduk (Katco) 2439.3 2661 3558 both
Moinkum (Northern, Katco) 889.1 1000
Southern Moinkum (Taukent/GRK) 442.5 1129 both
Kanzhugan (Taukent/GRK) 561.9
Chu-Sarysu,
Northern
Uvanas (Stepnoye-RU/GRK) 300.3 1192 both
Eastern Mynkuduk (Stepnoye-
RU/GRK)
1029.2
Central Mynkuduk (Ken Dala.kz) 1242.4 1800
Western Mynkuduk (Appak) 442.2 998
Inkai-1, 2, 3 (Inkai) 1636.7 1701 2047
Inkai-4 (South Inkai) 1701.4 1870 2030
Akdala (Betpak Dala) 1027.1 1095 1020
Budyonovskoye 1, 3 (Akbastau) 739.6 1203 1499
Budyonovskoye 2 (Karatau) 1708.4 2135 2115
Syrdarya,
Western
North and South Karamurun (GRK) 1016.7 1000
Irkol (Semizbai-U) 750 750
Kharasan 1 (Kyzylkum) 260.1 583 752
Kharasan 2 (Baiken-U) 262.2 603 888
Syrdarya,
Southern
Zarechnoye (Zarechnoye) 778.2 942 931
Northern,
Akmola
region
Semizbay (Semizbai-U) 224 411
RU-1 (Vostok, Zvezdnoye) 352.1 331
Total 17,803.4 21,317 22,451
163Introduction to uranium in situ recovery technology
be enriched to either natural uranium in the form of UF6 or LEU or blended with
HEU to create LEU.2 Plutonium and recycled uranium must also be recovered via
reprocessing spent fuel and then processed into fuel for reactors.
The difference between the second and third rows is one of reactor usage. Only
certain reactors can use mixed oxide (MOX) fuel that contains recycled plutonium
and depleted uranium. The same is true for reprocessed uranium (RepU), which
must be further enriched, yielding enriched reprocessed uranium (ERU), or blended
with HEU or medium enriched uranium (MEU) to achieve a suitable reactivity.
Another way of looking at this is that there is no broad commercial market for these
types of fuel. Thus, the uranium or uranium derivative not only needs to be pro-
cessed further (reprocessed), but it also needs a host reactor that can use it.
Note that it is possible to further subdivide the upper left-hand box, as the mate-
rial in this box is not all equally available to the market for a variety of reasons:
● Pipeline (in-process) uranium held by utilities● Strategic uranium inventory held by utilities● Inventory being held by commercial banks as collateral● Inventory being held by investment funds for exposure to uranium prices● Inventory being held by fuel banks in case of supply disruptions.
The arrows show the steps that uranium or uranium equivalents must go through
to be available to meet reactor needs. In this respect, governments might have some
natural uranium they are looking to place in the commercial market. This uranium
is not usually immediately available because the government must decide to release
Commercial
U3O8, natural and
enriched UF6
Tails material, Pu,
RepU, and
fabricated fuel
HEU, Pu, and
tails material
MOX and ERU
U3O8, natural and
enriched UF6
Government
Figure 8.1 Secondary supplies.
Source: Original by author.
2Normal uranium is uranium that has 235U content of 0.7%, the fissile content of uranium in nature.
216 Uranium for Nuclear Power
it, and in some cases this is subject to a public review. It is also the case that
governments and international organizations have created fuel banks that hold
inventories of LEU for potential use by utilities if certain conditions are met. These
inventories properly belong in the upper right-hand box because they are not avail-
able to the general market and would only be available to specific utilities/countries
under special circumstances.
As stated previously, secondary supplies in the second row require further pro-
cessing. From a practical standpoint, much of these supplies fall in the right-hand
column in that they are owned by governments. This is certainly true of all HEU
and much of the tails material.3 Thus, these supplies must be declared excess by
governments and processed further to get them in a commercially usable form. This
is not easy or straightforward, and when it comes to HEU, it can represent a hercu-
lean task. Even when it comes to tails material, there must be sufficient economic
enrichment capacity to enrich the tails.
Thus, not all secondary supplies are equal, and their availability is conditional
on their form and ownership. With this framework in mind, we will examine the
different types of secondary supplies and some transformations in enrichment and
reactor technology necessary to realize the full potential of these supplies.
8.1.2 The role of technology
In examining secondary supplies, it is important to understand the role of technol-
ogy. Technology impacts the accessibility of secondary supplies in a variety of
ways. Accessibility of supplies not only relates to any further processing that is nec-
essary, but the ability of reactors to use different forms of supply; in both of these
areas, technology can have a large impact on the role of secondary supplies for
nuclear power.
For example, the necessary technology needed to be in place to convert and
blend down HEU supplies so that they were available for reactor use. The same is
true for the enrichment of tails material. Although this technology does exist (tails
material is reenriched today), advances in enrichment technology can expand the
amount of tails material and hence secondary supplies that are economically avail-
able. If laser enrichment can achieve both superior economics and can enhance the
capability of enriching to lower tails assays, it will be able to transform more tails
into usable material and reduce the need for newly mined uranium. Enrichment
technology can also impact the amount of uranium available from underfeeding, the
situation in which enrichment plants operate at a lower average tails assay than
the one on which deliveries of uranium feed to the plants is predicated. However,
the primary effect here is to reduce the demand for new uranium production, as
lowering the tails assay reduces the amount of uranium needed in the production of
nuclear fuel.
3The tails controlled by Russian and Chinese enrichers are considered government owned, due to the
nature of these economies. Also, the DOE owns the tails inventories held in the US.
217Nuclear fuel from secondary supplies of uranium and plutonium
There is also the reactor dimension. Reactor fuel loadings must be designed (and
licensed) to use mixed oxide fuel and recycled uranium. If reactors using this type of
fuel are shut down and not replaced with others set up to utilize these fuel sources,
then the contribution of MOX and RepU will decline. Importantly, advances in reac-
tors in the future could allow reactors a greater use of recycled products or even
enable them to create their own fuel, such as in the case of fast breeder reactors.
Thus, the availability and usability of secondary supplies can change over time with
advances in technology to either produce or use secondary supplies.
8.2 Commercial inventories of naturaland enriched uranium
While commercial inventories of natural and enriched uranium are large, perhaps
on the order of 300,000 tU or more, they are all not readily available to displace
uranium production.4 Some of these inventories are in process (in the pipeline), ear-
marked to meet reactors needs. Still others are held for strategic reasons, to be
made available if there is a supply disruption or some other development. Suppliers
and utilities hold inventories for this reason. In addition, uranium funds and fuel
banks also hold inventories for very specific reasons.
Thus, there is a specific demand for inventories in addition to that associated with
the ongoing fuel needs of reactors. This demand depends on the overall need for ura-
nium and various reasons for holding it, and can be expected to vary over time as
reactor requirements change, the market is deemed more or less secure, and prices
fluctuate. This inventory demand itself can be quite large, and may be on the order
of 200,000 tU based on current and expected reactor requirements and market condi-
tions. Inventory supplies available to meet reactor requirements are those in excess
of these various needs. However, even though entities may be holding inventories in
excess of their needs, it does not mean these supplies are readily available to be sold
into market, as there may be various factors restricting their immediate sale or use.
Next is information on inventories held in key regions of the world.
8.2.1 United States
Data on inventories held by US companies is collected and reported by the US Energy
Information Administration (EIA). In its preliminary results for 2014, EIA reports that
US companies held almost 52,000 tU (135 million pounds U3O8 equivalent) in the
form of uranium concentrates (U3O8), natural UF6, enriched UF6, and fabricated fuel.5
Of this amount, 44,600 tU (116 million pounds U3O8) were held by utilities and
7300 tU (19 million pounds U3O8) were held by suppliers. Only a small percentage of
4This figure does not include uranium contained in tails material and HEU scheduled to be blended
down, reprocessed uranium, and the uranium equivalent of plutonium stocks.5U.S. Energy Information Administration, 2014 Uranium Marketing Annual Report, May 2015.
218 Uranium for Nuclear Power
these inventories are thought to be excess and become available for sale into the mar-
ket, although some of these inventories may have been purchased to meet future utility
needs and thus will be drawn down internally by utilities at a later date.
8.2.1.1 US government inventories
The US government also holds inventories earmarked for commercial use (or poten-
tial commercial use). Inventories of natural uranium acquired for past military pro-
grams have largely been disposed of. In addition, the US Department of Energy
(DOE) purchased some Russian HEU feed as part of the Russia-US HEU deal which
it has been disposing of. More recently, some DOE tails material has been enriched
and sold to US utilities. Most of DOE’s remaining inventories are currently in the
form of HEU or tails material, and will be discussed later in the chapter.
DOE’s sales and transfers of uranium inventories are subject to existing legisla-
tion, where the DOE Secretary must make a determination of “no adverse material
impact” to the domestic (US) nuclear fuel industry.6 For a number of years, the
DOE Secretary made this determination despite falling uranium and enrichment
prices, most notably in the wake of the Fukushima accident. The US nuclear fuel
industry objected to these determinations, and pushed for a set limit on how much
DOE could sell or transfer as a percentage of US uranium requirements.
In 2014, ConverDyn, a US uranium convertor, filed a lawsuit against DOE
relating to DOE sales and transfers of uranium alleging, among other things, that
the transfer of inventories was in violation of the USEC Privatization Act.
Largely in reaction to this lawsuit, DOE slightly reduced the amount of its
planned transfers, as shown in Table 8.1. Because it took this action, DOE
Table 8.1 Transfer volumes for Portsmouthcleanup & HEU down-blending in 2015 SD
Concentrates
(tU)
Conversion
services
(tU as UF6)
Enrichment services
(SWU)
2015 2500 2500 520,000
2016 2100 2100 520,000
2017 2100 2100 520,000
2018 2100 2100 520,000
2019 2100 2100 520,000
2020 992 992 520,000
2021 500 500 520,000
2022 500 500 520,000
2023 500 500 520,000
2024 500 500 520,000
Source: US Department of Energy (public information).
6The governing legislation is the USEC Privatization Act.
219Nuclear fuel from secondary supplies of uranium and plutonium
considered the matter settled and petitioned the court to end the case, but
ConverDyn continued its challenge.
In May 2015, there was legislation introduced in the US Congress to restrict the
ability of DOE to transfer uranium. This legislation, known as the Excess Uranium
Transparency and Accountability Act, would restrict DOE transfers to 2100 tU in
2016�2023 and 2700 tU in 2024 and beyond. The lawsuit and the proposed legisla-
tion underscore the fact that DOE can face some constraints when making decisions
to liquidate inventories.
8.2.2 Europe
The Euratom Supply Agency (ESA) reports inventories held by European Union
(EU) nuclear utilities, both strategic and pipeline. In its report for 2014 activity, the
ESA stated that EU utilities held 52,898 tU of uranium inventories, which it notes
covers up to 3 years of EU gross reactor requirements.7 The ESA advocates holding
inventories for security of supply, so not much of this inventory is likely surplus to
needs and thus would not available to the market. However, to the extent that some
EU reactors are shut down early as countries move away from nuclear power, it is
likely that some of this inventory would be drawn down to meet reactor needs or
sold, as there would be no reason to hold strategic or pipeline inventories for reac-
tors that will no longer be operating.
8.2.3 Japan
Typically, Japanese utilities have opted to hold rather large uranium inventories, as
Japan has no indigenous supplies of uranium. Thus, they were holding very large
inventories at the time of the Fukushima accident and these inventories have contin-
ued to grow as all Japanese reactors were shut down and remain offline, whereas
fuel deliveries under some contracts continued.8 UxC estimates that Japanese utili-
ties held almost 75,000 tU in 2014 based on an analysis of information available in
annual reports and other industry data.
A large share of these inventories are likely in excess of Japan’s inventory needs
as Japan will have fewer reactors operating in the future than prior to Fukushima.
How much is excess depends on how many reactors eventually come back online,
but most recently the Japanese government was targeting nuclear power to meet
20�22% of Japan’s electricity needs by 2030, down from approximately 30% prior
to the Fukushima accident.9 Another reason that Japanese utilities may want to hold
less inventory in the future is that its utility industry is undergoing deregulation and
this could impact the economics of holding inventory. To the extent that Japanese
utilities use excess inventories internally, it is likely that they will not need to buy
much uranium on the market for a relatively long time. In addition, it appears that
some of these inventories are being sold on the market or at least sold back to sup-
pliers which can deliver this material into other contracts.
8.2.4 China
Based on comparing China’s uranium imports and domestic production versus its
reactor needs, it is estimated that China has accumulated 105,000 tU of uranium
inventory.10 This is a huge amount, about twice the level of commercial inventory
holding in the United States, which has the world’s largest nuclear power program.
China is in the process of rapidly expanding its overall nuclear power capacity and
could overtake the US as the world’s largest program by the next decade, and this
prospective growth has likely motivated this large inventory buildup.
It is difficult to say how much, if any, of this inventory is excess to China’s
needs. Uranium is seen as having geopolitical importance in China, and because
China has little in the way of domestic uranium resources and it will take time for
its investments in foreign uranium mines to pay off, building large inventories rela-
tively quickly is the best way to assure the availability of uranium supplies. It may
be the case that some of these inventories are worked off internally as China’s for-
eign uranium mines come into production and its domestic enrichment production
capacity expands. Once China has more uranium and enrichment production in
place, it may be more comfortable holding less material in inventory for strategic
reasons.
A critical question here is China’s nuclear power growth. If China greatly
expands its nuclear power capacity, it is less likely to reduce its inventory holdings
and may even add further to its inventory. However, if China’s nuclear power
growth stagnates, it may look to reduce its inventory holdings. Given the role of the
Chinese government in China’s nuclear power program and its economy in general,
it makes little sense to try and differentiate inventories controlled by commercial
entities inside of China and inventories controlled by the government.
8.2.5 Russia
Russia is also thought to hold large uranium inventories, as it does not have large,
economic uranium resources. However, inventories held in Russia are a state secret,
and thus there is no transparency with respect to inventory holding or disposition
plans as is the case in the United States. In the past, the Soviet Union held large
inventories of uranium based on deliveries from uranium producing regions like
Kazakhstan and Uzbekistan and Eastern Europe (East Germany). These were used
to fuel its growing nuclear power program which received a serious setback follow-
ing the Chernobyl accident.
10This is an UxC estimate. Note that this amount would be in addition to any inventory that the Chinese
government was holding for civil or military purposes prior to its importing of uranium.
221Nuclear fuel from secondary supplies of uranium and plutonium
It is thought that while Russia still holds notable stocks of natural uranium and
UF6, some of this from repatriated HEU feed, it has utilized much of this material
over time. As a result, it may be the case that a larger portion of Russian inventories
and thus potential secondary supplies are currently in the form of tails material and
HEU, as well as being generated via underfeeding. As will be discussed later in this
chapter, Russia plans to use these inventories to meet domestic demand and export
commitments. Like the case with China, it makes little sense to differentiate inven-
tory holding between government and commercial entities as the Russian govern-
ment is heavily involved in all phases of Russian nuclear power and fuel cycle.
8.3 Other natural and enriched inventories
Inventories are held by other entities for financial or supply assurance and nonpro-
liferation reasons.
8.3.1 Uranium participation corporation
In 2005, the Uranium Participation Corporation (UPC) was created as a publicly
traded stock company listed on the Toronto Exchange as a way of giving investors
the opportunity to participate in the then rapid appreciation of uranium prices. UPC
purchases uranium (in the form of either U3O8 or UF6) in the market and investors
buy shares in UPC to underwrite these purchases and to own a share of UPC’s ura-
nium holdings. UPC currently holds a little more than 5500 tU (14.5 million pounds
U3O8) equivalent in inventory in the form of U3O8 and UF6.11 About two-thirds of
this inventory is held in the form of U3O8.
This type of inventory holding is solely driven by financial considerations to
give investors exposure to the uranium price and market. So far, the fund has been
operated in a way that there have only been purchases into the fund, with no sales
from the fund. However, UPC has loaned out some of its inventory to defray some
of its holding costs and has sold off some of the conversion component of the UF6it holds. UPC can sell any or all of its inventory at anytime (if it could not the
inventory and the related stock would have no value). One potential exit strategy
for UPC is to sell the entire inventory to one buyer.
8.3.2 Fuel banks
Another type of inventory that is theoretically available to meet reactor needs is
that held in fuel banks. Fuel banks were established to provide a backup to the com-
mercial market in case of supply disruptions. Access to supplies in fuel banks is
generally conditioned on a supply disruption that cannot be remedied by the market
and the receiving party having good nonproliferation credentials. While there is no
11 “Uranium Participation Corporation Reports Net Asset Value at April 30, 2015,” May 5, 2015.
222 Uranium for Nuclear Power
requirement that countries forgo the right to develop their own fuel cycle facilities
(including enrichment and reprocessing) to have access to fuel bank material, the
aim of fuel banks is to provide fuel supply assurances such that development of
these facilities may be deemed unnecessary.
There are currently two fuel banks, one in Russia and one in the United States.
In addition, the International Atomic Energy Agency (IAEA) is in the process of
developing a fuel bank, and the United States has instituted a MOX backup inven-
tory program, which while different in function from the other fuel banks, holds
inventories of LEU in case of disruptions/problems with supplies from commercial
contracts signed for MOX supplies created in US facilities.12 Table 8.2 next sum-
marizes the various nuclear fuel banks.
The Russian fuel bank, which was set up under IAEA auspices, holds 120 t of
LEU. The EUP for the IAEA fuel bank in Kazakhstan has not yet been purchased,
although funds have been allocated. The irony is that the funds allocated for the
fuel purchases can now be used to acquire considerably more EUP than when they
were first committed because enrichment and uranium prices have fallen dramati-
cally in recent years.13 The US fuel bank, the American Assured Fuel Supply
(AAFS) holds 230 t of LEU which is equivalent to about six reloads for a
1000 MWe reactor. Direct access to this fuel bank is limited to US companies, and
non-US companies may only have access to this material via their US suppliers.
The MOX Inventory Backup Program is estimated by UxC to contain about 168 t
of LEU and continues to grow to the extent that material from the US highly
enriched down-blending program is placed in this stockpile. It, too, would be lim-
ited to US companies, specifically utilities which would have contracts for MOX
fuel produced in the United States, if such a program ever moves forward.
Fuel banks create their own demand separate from reactor requirements and nor-
mal inventory holding. The supplies held in fuel banks are rather low relative to
overall reactor needs, representing less than 20 reloads for a 1000 MWe reactor in
total, most of which is located in the US. Furthermore, it may be that uranium from
fuel banks is used if certain situations arise, but in this case these inventories would
be replenished. The exception to this would be the uranium contained in the MOX
backup inventory program that would no longer be necessary if the United States
does not pursue a MOX program. Thus, for the most part, fuel banks do not repre-
sent a net future supply to the market, as any outflows to the market would need to
be replaced by purchases, unless it was decided that they were no longer necessary.
8.3.3 Future trends
Much of the current buildup and use of commercial uranium inventories is a vestige
of the excess inventory supplies created by the Fukushima accident, along with the
12 In June 2015, the IAEA Board of Governors approved agreements to establish a fuel bank in
Kazakhstan. The IAEA has not yet procured the EUP. It has funds to purchase 70�74 t of EUP at
today’s prices.13 Funds were initially obtained in 2010, when enrichment prices were about twice what they are today.
223Nuclear fuel from secondary supplies of uranium and plutonium
Table 8.2 Nuclear fuel banks
Russian fuel bank (under
IAEA auspices)
IAEA fuel bank AAFS (US) MOX backup inventory
program (US)
Status Inaugurated in December
2010
Establishment approved in
2010; siting agreements
signed in 2015
Availability announced in
August 2011
Some material available,
will be fully formed by
the end of 2015
Location Angarsk electrolysis chemical
combine, (International
Uranium Enrichment
Center), Russia
Ulba Metallurgical Plant,
Kazakhstan
United States United States
Ownership while
in storage
Russia IAEA United States United States
Material available 120 t LEU Financial funds: $125
million and h25 million
230 t LEU B168 t LEU
Storage costs Russia IAEA United States United States
desire of the Russian and US governments to dispose of certain inventories stem-
ming from past military programs, or, in the case of Russia, utilize excess enrich-
ment capacity from past military programs. In general, more inventories are being
held than desired, and these inventories will be pared down over time as reactor
requirements grow and utilities are better able to consume them. This may take
some time, as witnessed by the difficulty that Japan is encountering in restarting its
reactors. An exception to this could be China, which may want to continue to build
inventories for strategic reasons, although this would be a function of its underlying
nuclear capacity growth and development of its foreign uranium projects.
US government inventories available for commercial nuclear use are winding
down, but their disposal may be even more protracted depending on how a current
lawsuit and legislative initiatives play out. It also depends if DOE is able to enrich
its tails material in the future, as its tails inventories need further processing.
However, the enrichment of tails would occur over a number of years as there
would be limits to how much could be processed in any one year.
In theory, inventories held by UPC and various fuel banks could be held indef-
initely and thus may never be used. These inventories could build over time,
depending on market and geopolitical events. For instance, if investors thought
that uranium prices would appreciate notably, there may be more demand to gain
exposure to uranium through UPC’s holdings. There could be more fuel banks or
fuel bank type arrangements as well. Recently, at the ATOMEXPO 2015 confer-
ence in Russia, it was announced that the BRICS states (Brazil, Russia, India,
China, and South Africa) were considering setting up their own nuclear fuel
reserve. In general, these types of arrangements create more demand for nuclear
fuel; although they are potential sources of supply should certain developments
occur in the future.
8.4 HEU supplies
8.4.1 Background
A large source of secondary uranium supplies has been HEU from dismantled
nuclear warheads that have been blended down for reactor use. The enrichment
level for nuclear weapons is typically over 90w/o. HEU can also come from fuel for
nuclear-powered submarines, although this is not as large a source of HEU as weap-
ons material.14
The notion of blending down HEU for reactor fuel is relatively new, and the
prospect for doing this was first mentioned in a 1989 article by Combs and Neff
that explored unconventional possibilities for nuclear fuel trade between the United
14HEU is defined as uranium enriched to over 20w/o235U, and can include material from nuclear weap-
ons, submarine fuel, and fuel for certain research reactors. Weapons material has enrichment levels in
excess of 90w/o.
225Nuclear fuel from secondary supplies of uranium and plutonium
States and the Soviet Union.15 Neff went on to propose what later became the
Russian�US HEU deal (discussed next) in an op-ed in The New York Times.16
For HEU from nuclear weapons to be transformed into reactor fuel, it has to
undergo a number of steps, one it is removed from the warhead itself. It must be
converted (in the case of Russia it was converted from a metal to highly enriched
UF6 gas) and be blended down to reactor grade (less than 5w/o235U). HEU can be
blended down using natural uranium, tails material, or slightly enriched uranium.
The 235U content of the blendstock needs to be below the assay of the final product,
and the closer the assay of the blendstock is to that of the final product, the more
blendstock that needs to be utilized to get to the desired product assay.
8.4.2 The Russian�US HEU deal
One of the major sources of secondary supplies in recent years has been the HEU
deal between Russia and the United States, which involved the delivery of
150,000 tU of uranium feed and 90 million SWU contained in about 14,400 t of
LEU delivered over a 20-year period that ended in 2013.17 As noted previously, the
HEU deal was suggested by Neff and eventually led to an agreement between the
Russian and US governments in 1993. The deal involved blending down 500 t of
HEU, equivalent to the content of 20,000 nuclear warheads, into reactor-grade fuel.
This government-to-government agreement was followed by an implementing con-
tract between the United States Enrichment Corporation (USEC), acting as
the executive agent for the United States, and Techsnabexport (TENEX) acting as
the executive agent for Russia. USEC purchased the enrichment component of the
down-blended LEU (about 90 million SWU), but not the feed component. The
inability of the feed component to be sold at first threatened to scuttle the deal,
but the US DOE stepped in to purchase some of the feed, and later a consortium of
the uranium producers AREVA and Cameco, and the trader NUKEM, entered into
an agreement to purchase much of the remaining feed, although a portion
(38,000 tU) was retained by TENEX and returned to Russia.
The blendstock in the Russian�US HEU deal was slightly enriched tails mate-
rial. The reason that this blendstock was used is because HEU has a relatively
high concentration of 234U and 236U, while tails have a much lower proportion of
these isotopes, both a function of the enrichment process. Blending HEU with a
blendstock that has a lower concentration of these undesirable isotopes makes
it possible to achieve a low enough concentration of these undesirable isotopes
in the final product to meet American Society for Testing and Materials (ASTM)
specifications.
15Combs, Jeff and Dr Thomas Neff, “The Soviets, SWU, and U—A Win-Win Solution?” NYNCO
Newsletter, November 1989.16Dr Thomas Neff, “A Grand Uranium Bargain,” Op-ed, The New York Times, October 24, 1991.17 Some of this discussion comes from the WNA’s “Military Warheads as a Source of Nuclear Fuel”
updated August 2014.
226 Uranium for Nuclear Power
Bukharin notes that in the US-Russia HEU deal 8555 t of depleted uranium (tails)
at an assay of 0.25w/o were enriched using 5.34 million SWU to create 916.6 tU of
LEU at a 1.5w/o product assay. This was blended with 30 t of HEU to produce
946.6 tU of LEU at a 4.4% product assay containing 5.52 million SWU and 9000 tU
(about 24 million pounds U3O8).18 It is noteworthy that almost as much enrichment
was used to create the blendstock as was contained in the final product.
8.4.3 US HEU blend down
The United States has also blended down some of its HEU, but not nearly to the
extent as Russia. The US initially declared 174 t of HEU excess to its needs and
later added to this amount, although not all of this material has become available
for down-blending. Some of it was utilized for naval propulsion and a portion was
declared waste, reducing the total volumes available to DOE for down-blending
into reactor-grade fuel. Note that US stocks of HEU are reported to be lower than
those of Russia and some of these are contaminated, making them unsuitable to be
easily converted to reactor fuel.
In 1997, DOE signed a memorandum of understanding with the Tennessee
Valley Authority (TVA), to use off-spec LEU in TVA’s reactors; after a demonstra-
tion was completed, DOE and TVA entered into an agreement in 2001. To create
this fuel, DOE agreed to supply approximately 33 t of off-spec HEU to be blended
down with natural uranium to reduce the 236U content (along with other impurities)
and make the resulting product more suitable for reactor use. This program is
known as the blend low-enriched uranium (BLEU) program. In 2008, additional
HEU was added to this program. To date, a total of 46 t of HEU has been processed
as part of this program and it has been extended by down-blending an additional
small quantity of off-spec HEU to be loaded between 2017 and 2023.
What has not gone into the BLEU program has largely gone into the AAFS fuel
bank and the MOX backup inventory program discussed earlier. Less than 20 t of
HEU remains to be blended down. Overall, DOE anticipates completing the down-
blending of the available HEU by 2024 at an expected annual rate of 2�3 tU.
A portion of the down-blended HEU has been used as payment to the companies
providing the down-blending service (payment in kind) and has been subsequently
sold into the market.
8.4.4 Future blend down of HEU
At the moment, there appears to be little prospect of another Russian�US deal,
despite the great success of the deal that did take place and the fact that there are
still large stocks of HEU. Importantly, Russia is not in the dire economic straits it
was in during the 1990s and US and Russian relations have currently deteriorated
quite a bit due to the situation in Ukraine.
18Oleg Bukharin, “Understanding Russia’s Uranium Enrichment Complex,” Science and Global
Security, 12: 193�218, 2004.
227Nuclear fuel from secondary supplies of uranium and plutonium
At the same time, it is possible that Russia could choose to blend down addi-
tional quantities of HEU for its domestic use or perhaps for its export market.
Overall, the USSR produced an estimated 1200 t of HEU, of which 500 t were used
for the HEU Agreement, leaving B700 t of HEU in Russia.19 The Russian govern-
ment has no formal process to declaring the material surplus and has given no indi-
cation at this time that it would consider doing so. It is interesting to note that, for
internal use, Russia can blend its HEU with depleted uranium (tails) as it does not
have to worry about meeting ASTM specifications for the final product. This
approach is more economic as it preserves both Russian natural uranium supply and
its enrichment capacity.
As mentioned previously, the United States is still planning to blend down some
additional HEU, perhaps at the rate of 2�3 t per year, but this effort has been sty-
mied recently by the decline in nuclear fuel market prices following the Fukushima
accident, making the uranium and enrichment components of the HEU less valuable
and less able to cover the costs of blending down the HEU. Because the United
States no longer has enrichment capacity based on US technology, it cannot cur-
rently create more HEU and thus does not have SWU capacity based on US tech-
nology to enrich uranium that can be used in the production of tritium, which is
used a booster in thermonuclear weapons.20 This can be a motivation for the US to
preserve its inventory of HEU, as it can be blended down and used for tritium pro-
duction in the future.
8.5 Recycled uranium and plutonium21
Uranium and plutonium derived from reprocessing spent fuel can be used as fuel in
light water reactors. The issue here is that when the original nuclear fuel is
“burned” in a reactor, it is not completely consumed, but most of the uranium can
be recovered along with some plutonium, which is created as part of the fission pro-
cess. Uranium recovered from spent fuel after reprocessing (RepU) needs to be fur-
ther enriched yielding ERU, but substitutes for freshly mined uranium.22 Plutonium
is mixed with depleted uranium (tails) to create a MOX fuel.
19The International Panel on Fissile Materials estimates that Russia had 6956 120 t of HEU at the end
of 2012. International Panel on Fissile Materials, Global Fissile Material Report 2013.20The only current enrichment technology in the United States that is commercially operable is that
owned and operated by URENCO, a British/Dutch/German consortium. There are usually prohibitions
against using non-U.S. enrichment technology and uranium for military purposes in the U.S., and in
any case, the U.S. has limited itself to the use of U.S. enrichment technology and U.S.-origin uranium
feed for these ends. The U.S. DOE has been funding the American Centrifuge Program in large part
because it might need U.S. domestic technology in the future to satisfy some of its military programs.21Much of this discussion is based on an essay entitled “The Estimated Impact of Recycling on the U3O8
Market,” UxC Uranium Market Outlook Q1 2015, June 2015.22RepU has about a 1w/o fissile content, which is greater than uranium found in nature (0.7w/o), but
RepU must be enriched to higher levels than natural uranium to yield the same performance as freshly
mined uranium due to the presence of undesirable isotopes. It also can be blended with HEU or MEU
to achieve these same results.
228 Uranium for Nuclear Power
Recycle can be done for economic reasons but is also done to reduce stocks of
plutonium that have accumulated over time. It is a step that necessarily follows
reprocessing of spent fuel, which is undertaken both to facilitate the management of
spent fuel as well as to recover its fissile products and byproducts for further use as
fuel. The economics of recycle, which a function of the costs of reprocessing and
MOX fuel fabrication as well as market prices for uranium enrichment, were not
very attractive prior to Fukushima, but after the accident and the subsequent decline
in nuclear fuel demand and prices, recycle economics became even more
problematic.23 Whereas in 2007 some market participants were worried that the
world was “running out” of uranium, there is now much less concern about the
availability of future uranium supplies.
Current estimates of civil separated Pu place the Pu inventory at 260 t based on
data reported to the IAEA. In addition, plutonium originally created for military
purposes can be used as fuel. After the Cold War, more than 100 t of military
Pu were declared excess by the US and Russia. In July 2011, the United States and
Russia each agreed to dispose of 34 t of weapons plutonium by converting
this material into MOX fuel for use in nuclear reactors.24 While current inventories
of RepU are not publicly available, UxC estimates these to be 75,000�90,000 t
based on past reprocessing activity.
Looking at the future, UxC estimates that nearly 45,500 t will be reprocessed
over the 2014�2030 period, yielding about 450 t of separated Pu and about
43,000 t of RepU. To the extent that the resulting RepU and plutonium are recycled
in reactors, they will displace the need for newly mined uranium. For instance, the
ESA notes that the use of MOX fuel resulted in a savings of 1156 tU in 2014.25
UxC estimates that MOX will displace about 16,500 tU (43 million pounds U3O8)
over the 2014�2030 period, equaling about 1.2% of the almost 1.4 billion tU (3.566
billion pounds U3O8) projected to be required over the 17-year period. As far as ERU
is concerned the reduction will be 20,270 tU (52.7 million pounds U3O8) over the
same period equaling about 1.5% of the uranium required over this period. Thus, in
total, MOX and ERU fuel are projected to displace about 36,800 tU (almost 100 mil-
lion pounds U3O8) over the 2014�2030 period, accounting for about 2.7% of the ura-
nium required to fuel reactors over this period, as shown in Fig. 8.2.
8.6 Future of recycled uranium and plutonium
As shown in Fig. 8.2, current plans are for the contribution of recycled uranium
and plutonium to decline over time. A major reason for this is that some of the
reactors that use MOX and RepU fuel will be shut down. The use of recycled
23When it comes to reenriching RepU, there is an added cost associated with dedicating a portion of an
enrichment plant’s capacity for this purpose to avoid contamination of centrifuges with undesirable
isotopes.24 International Panel on Fissile Materials, Global Fissile Material Report 2013.25EURATOM Supply Agency Annual Report 2014, p. 29.
229Nuclear fuel from secondary supplies of uranium and plutonium
products beyond 2030 will be a function of the future economics of uranium and
enrichment production, future efforts to reduce plutonium stocks (both civilian
and military), future operation of reactors that use these products, and future
advances in reactors and reactor fuel design. This last point will be examined later
in this chapter.
8.7 Enrichment of tails material
8.7.1 Overview
As discussed in the introduction, the fissile content of uranium that is mined is not
completely consumed when it is processed or burned in a reactor. When uranium is
enriched, some of the fissile content remains in the waste stream, or tails. The assay
of the tails depends on whether more or less uranium (and less or more enrichment)
was used to create the enriched product at the time of enrichment. Tails assays can
generally range from 0.10w/o to 0.40w/o and can fall outside of this range under
extreme situations. Because the 235U assay of uranium found in nature is 0.711w/o,
tails with an assay of 0.40w/o have much of the fissile content remaining, while tails
Figure 8.2 Combined annual uranium displacement impact of MOX and ERU Fuel.
Source: The Ux Consulting Company, LLC.
230 Uranium for Nuclear Power
as low as 0.10w/o have little fissile content remaining, and thus are not economic to
reenrich. The economics of enriching tails material depends both on the assay of
the tails and the cost of enrichment, as well as the value of uranium feed. In the
case of centrifuge technology, the cost of enrichment can be quite low as variable
costs (chiefly the consumption of electricity) are low and enrichers generally prefer
to keep centrifuges spinning because shutting them down and restarting can lead to
failures and loss of capacity.
All enrichers can reenrich tails, but to date, the vast bulk of tails reenrichment
has been done in Russia, where Russia has enriched not only domestic tails mate-
rial, but also tails imported from Europe. The United States has had some well-
publicized tails reenrichment activity, but its scope has been far less than that of
Russia, although it could be expanded with advances in enrichment technology.
8.7.2 Russian tails reenrichment activities
Russia has represented by far the largest source of tails reenrichment, due to its
large and relatively inexpensive enrichment capacity (developed in large part to
meet earlier military needs), early transition to centrifuge technology, its histori-
cally large stock of tails, and its relatively low level of natural uranium resources.
8.7.2.1 Domestic consumption and export
The Chernobyl accident and the end of the Cold War left Russia with a consider-
able amount of excess enrichment production capacity. Furthermore, following the
breakup of the Soviet Union, the major uranium producing regions of Kazakhstan
and Uzbekistan were no longer captive to Russia. Because Russia does not have
large economic uranium resources, it made economic sense for it to enrich tails
material using its excess enrichment capacity and its abundant supply of tails to
fuel its domestic reactors and for export. Bukharin notes that in 1992, Minatom (the
then Russian Ministry of Atomic Energy) planned to increase the enrichment capac-
ity devoted to tails enrichment from 1.29 million SWU in 1993 to 6.44 million
SWU in 2000 through 2010.26 Also, the Russian enrichment complex was originally
supposed to enrich low assay tails and move to higher assay tails, but it appears
that it did not follow this program as it was enriching 0.36w/o (high assay) tails as
early as 1992.27
It is thought that by now Russia has enriched most or all of its high assay tails,
and currently enriches tails material with average assays of around 0.20w/o. Russia
does have considerable enrichment capacity in excess of the demand for its enrich-
ment services, and is thought to operate this capacity at a tails assay of 0.10w/o to
further strip its available tails material. It is estimated that Russia currently uses
one-third of its excess SWU capacity to enrich tails. This could produce on the
26Bukharin, op. cit.27Bukharin, op. cit.
231Nuclear fuel from secondary supplies of uranium and plutonium
order of 1153 tU (3 million pounds U3O8) per year to the extent that this tails feed-
stock and SWU capacity are available in the future.28
8.7.2.2 HEU blendstock
As noted previously, uranium tails enriched to an assay of 1.5w/o were used as the
blendstock for HEU in the Russian�US HEU deal. This process consumed a little
over 5 million SWU of Russia’s enrichment capacity on an annual basis. While the
enrichment component of the LEU derived from the HEU was sold to USEC, a por-
tion of the uranium feed was returned to Russia and the remainder was purchased
by a Western consortium for sale in the market. The material returned to Russia has
become part of the state-owned inventory and is not readily available to the
Russian nuclear industry at this time.
8.7.2.3 AREVA and URENCO tails
In the late 1990s, Russia entered into agreements to enrich tails material from AREVA
and URENCO. The tails shipped to Russia were relatively high assay (0.30w/o and
above). While some of the tails were likely used for the HEU blendstock, Russia
returned both natural (0.711w/o) uranium and enriched uranium to the European enri-
chers. Overall, these deals represented a sizeable amount of natural uranium supply,
over 2300 tU (6 million pounds U3O8) per year that were returned to the European
enrichers plus another 923 tU (2.4 million pounds U3O8) contained in EUP. With the
contracts expiring in 2009 and 2010, Russia decided it would not extend the contracts
as they were a public relations nightmare and also they were economically unjustified
in light of the then new conditions in the enrichment market.
Russia continues to enrich tails material with its excess SWU capacity, as it
looks to utilize its excess enrichment capacity to generate additional uranium sup-
plies. The Fukushima accident left Russia with considerable excess SWU capacity
and the enrichment market oversupplied in general. As a consequence of these
developments, Russia is likely to have ample SWU capacity to devote to the enrich-
ment of tails for the foreseeable future. One issue that may confront Russia is the
lack of relatively high assay tails material to enrich.
8.7.3 US tails enrichment
Tails material has also been enriched to natural uranium levels in the United States.
There have been two tails deals in the United States, both involving Energy
Northwest (ENW) with the enrichment done at USEC’s then operating gaseous dif-
fusion plant in Paducah, Kentucky. The first, done in 2005, involved tails enriched
by USEC and supplied to ENW and involved about 1923 tU (5 million pounds
28Given the poor quality of Russian tails, UxC estimates that about one-third of Russia’s excess enrich-
ment capacity is devoted to tails enrichment, while the remaining two-thirds of capacity is utilized for
underfeeding. This analysis assumes stripping form 0.20w/o to 0.10w/o, the assumed operational tails
for Russia.
232 Uranium for Nuclear Power
U3O8). The second, from 2012, involved 9075 t of tails enriched by USEC to 4.4w/owith the product being delivered to ENW and TVA. This deal resulted in the pro-
duction and sale of 482 t of EUP and another 1600 tU of UF6.
Enrichment of tails in the United States was performed using gaseous diffusion
enrichment technology, which is more expensive than centrifuge but does not incur
the same operational penalty for being shut down or having its output cut back.
Thus, it is more sensitive to market price pressures but can react more definitively
toward them. In this regard, there was to have been a third tails enrichment cam-
paign in the US following the 1-year tails enrichment program of ENW/TVA/
USEC, but it became uneconomic when uranium prices fell following the
Fukushima accident. Given the relatively high cost of enrichment using the anti-
quated gaseous diffusion technology employed by USEC, the enrichment of tails
became uneconomic once the uranium price dropped below $60.
8.7.4 Future of enriching tails material
As previously noted, Russia is likely to have excess enrichment capacity for a long
time in the future, so it will clearly have the ability to reenrich tails; the question is
what assays of tails material it will have at its disposal to reenrich. In this respect,
if Russia is enriching at a 0.10w/o tails assay now, it will not be economic or even
technically feasible to further strip the resulting tails due to their very low fissile
content.
As mentioned previously, DOE is currently considering enriching tails at
Paducah using an enrichment plant based on the Silex laser technology that would
be built by GE-Hitachi Global Laser Enrichment (GLE). This technology promises
lower costs and an ability to strip tails to lower levels, making further tails enrich-
ment more economic. As of now, plans are to enrich tails to create 2000 tU (5.2
million pounds U3O8) of natural uranium per year for a period of 15 years, resulting
in 30,000 tU (78 million pounds U3O8) of additional secondary supplies via tails
enrichment.29 In addition to the economics of future tails enrichment in the US, the
ability of DOE to sell uranium into the market has been challenged and some
restrictions may be placed on the annual quantities of uranium that can be sold by
DOE, including those generated by the enrichment of tails, based on legislation that
has recently been proposed.
Further in the future, the enrichment of tails could represent an important source
of uranium feed to the extent that enrichment technology improves. There are two
facets of this improvement. First is the economics of enrichment, which depends on
advances in enrichment technology. The lower the cost of enrichment, the greater
the quantity of tails that can be economically reenriched. Second, also associated
with technology advances is the ability to physically enrich at lower tails assays,
which expands the recovery of uranium. Thus, if advances in technology occur
which both reduce costs and lower the assay at which the enrichment plant can
physically strip tails, then more tails material can be economically processed.
29The RFP for this work has a provision for two optional extensions of five years each.
233Nuclear fuel from secondary supplies of uranium and plutonium
Ultimately, there will be diminishing returns even with tremendous advances in
technology as the assays of the tails feedstock (the tails material being enriched)
would continuously decline as more of the fissile content of uranium was recovered
through the more efficient processes. However, this would likely take some time to
occur, as there is an abundance of tails material, especially at lower tails assays.
8.8 Underfeeding
Another type of secondary supply is created when enrichers operate at lower tails
assays than those at which they transact. The result of this activity, which is com-
monly referred to as underfeeding, is that enrichers end up with extra uranium feed
they sell into the market. This material ends up competing with newly mined ura-
nium and inventories in meeting the demand for uranium.
The amount of underfeeding that will take place in the future depends on both
the tails assay at which enrichers sell their services to utilities and the tails assay at
which they operate their plants. This, in turn, depends on the competitiveness of the
enrichment market and the overall need for enrichment. In the recent past, the aver-
age transaction tails assay has been notably above the average operating tails assay
and even well above the optimal tails assay, which has dipped below 0.20w/o.30 As
a result, enrichers have accumulated a lot of uranium and have been big sellers on
the spot, mid-term, and even long-term contract markets.
8.8.1 Underfeeding vs tails enrichment
Tails enrichment and underfeeding are sometimes thought of as essentially the
same thing—utilizing excess enrichment capacity to create uranium, but they are
different in important respects. The enrichment of tails creates new uranium supply
by enriching tails material (depleted uranium) up to the fissile level at which ura-
nium is found in nature. Underfeeding, on the other hand, reduces the amount of
uranium that is needed in the future by using less uranium in the enrichment pro-
cess to produce the same amount of enriched product. Uranium “created” as the
result of underfeeding can be thought of as a secondary supply (and perhaps a
major source of future secondary supply) in that it does not come directly from ura-
nium producers. However, it only becomes a supply if more uranium is delivered to
enrichers by utilities than is actually needed in the enrichment process; no new ura-
nium is created.
While there are inventories that exist today that have been created by underfeed-
ing, future supplies of uranium from underfeeding depend on the contracting prero-
gatives of enrichers and utilities, and thus do not relate to any current physical
supplies of uranium. That is, if all enrichers transacted enrichment services at the
30The optimal tails assay is the assay that provides the lowest-cost mix of uranium feed and enrichment
in the creation of enriched product. This valuation is usually based on the prevailing spot prices of
enrichment, uranium, and conversion.
234 Uranium for Nuclear Power
same tails assays at which they operated their plants, there no uranium would come
from underfeeding in the future.
The difference between underfeeding and tails enrichment can be looked at
another way as well. Underfeeding relates to the reduction of the future need for
uranium, while tails enrichment, in essence, reduces the past requirements for ura-
nium to the extent that tails from previous enrichment activities are still available to
be reenriched.
Finally, another difference is that enrichment plants can be built specifically to
enrich tails, and in this sense they can be thought of as being akin to uranium
mines. Indeed, the proposed enrichment plant to be built at Paducah by GLE with
laser enrichment technology developed originally by Silex Systems in Australia
would only be used for the enrichment of tails, and not for providing conventional
toll enrichment services.31 The plant would be configured a little differently for this
purpose.32 Other plants or cascades within existing centrifuge plants can be simi-
larly configured to enrich tails in the future.
8.8.2 Future underfeeding trends
Most recently, as competition has intensified in the enrichment market, it appears
that enrichers are giving their customers more flexibility in selecting their tails
assay, allowing them to choose lower tails assays, perhaps below 0.20w/o. To the
extent this flexibility exists and utilities opt to transact at lower tails, the amount of
uranium they need to deliver to enrichers will decline and thus the secondary sup-
plies generated from underfeeding will also decline. There will likely always be
some degree of underfeeding even if utilities are given complete flexibility in
selecting their tails because the optimal tails for enrichers to operate their plants
will be lower than the optimal tails at which utilities transact.33 Finally, declining
secondary supplies from underfeeding should not be taken as a sign that more ura-
nium needs to be produced. To the extent that secondary supplies from underfeed-
ing decline, so does the amount of uranium that utilities need to supply to enrichers
to meet their enrichment contract obligations. These two developments offset each
other. All that is occurring is that enriches are giving their utility customers a
greater ability to select tails and thus to optimize their fuel costs.
8.9 Implications of generation IV nuclear reactors
Advances in reactors can bring with them new fuel designs that need less, or per-
haps no, natural uranium at all to fuel them. In this case, some of the secondary
supplies discussed here, including plutonium and depleted uranium, can meet more
of future demand. However, this is more of a demand-side development than
31 Silex Systems announcements: www.silex.com.au32 Private communication: Michael Goldsworthy, Silex Systems.33This is because enrichers’ costs are typically lower than market prices of enrichment.
235Nuclear fuel from secondary supplies of uranium and plutonium
a supply-side one, because it depends mostly on what happens with respect to reac-
tor technology.
As an example of this type of consumption, Russia is planning to use weapons
plutonium manufactured into MOX fuel in its BN-600 and BN-800 fast neutron,
breeder reactors.34 Fast reactors can consume “nonstandard” secondary supplies
including plutonium and depleted uranium. In general, generation IV reactors can
use a wide variety of fuel, including thorium, and thus are less dependent on tradi-
tional uranium supplies.35
Advanced reactors and different fuel cycles are most likely to be sought by
countries that have low uranium resources. The chief candidates here are China,
India, and Russia, all of which possess relatively low levels of uranium resources
and production, but all of which have relatively large and expanding nuclear power
programs. China and Russia have active breeder programs. India, which has abun-
dant thorium reserves, is pursuing the goal of moving to a thorium cycle. However,
the development of these reactors is progressing slowly, and so far they have had
minimal impact on the consumption of secondary supplies and overall uranium
requirements.
8.9.1 Future trends
To the extent that the movement to advanced reactors is driven by the desire to econ-
omize on the future use of uranium, recent market developments have likely reduced
this incentive. In this respect, the need to economize on uranium and nuclear fuel in
general has declined following Fukushima and the lower demands and prices it
brought. Also, advances in enrichment technology (later generation centrifuges and
the potential of laser enrichment) should continue to reduce the need for uranium.
One question is whether advances in enrichment technology can outpace those in
reactor technology and which will have the greatest relative impact on uranium
demand. So far, it appears that enrichment technology is winning this battle.
8.10 Summary: The declining but continuing roleof secondary supplies
Fig. 8.3 shows the estimated historical use (from 2008) and projected future use
(through 2030) of secondary supplies relative to both uranium production from
existing and new facilities and a forecast range of future uranium requirements.36
The sources of secondary supplies include drawdown of excess commercial inven-
tories by utilities and suppliers, US government inventory sales and transfers, tails
34 Pavel Podvig, “Disposition of Excess Military Nuclear Material, February 2012.”35 Six types of generation IV reactors along with their fuel supplies, are discussed in the World Nuclear
Association’s “Generation IV Reactors” (updated June 2015).36This is based on UxC information, including its midcase uranium production forecast, which includes
projected production from existing and new mines.
236 Uranium for Nuclear Power
enrichment and underfeeding (particularly in Russia), ERU, and uranium equiva-
lents contained in MOX fuel. It also includes uranium deliveries from the
Russian�US HEU deal for the period prior to 2014.
On inspection of Fig. 8.3, it is clear that the contribution of secondary supplies
has declined recently, and is projected to decline further over time. The large drop-
off in secondary supplies after 2013 relates to the end of the HEU deal. From 2013
to 2030, the annual level of secondary supplies is essentially halved. Pretty much
all categories of secondary supplies fall off during this period, and supplies from
the drawdown/sale of currently excess utility inventories are projected to disappear
entirely by 2030. During this time, the excess inventories built up as a result of the
Fukushima accident are expected to be worked off. Of course, these projections are
based on current forecasts of uranium requirements, contracting practices of enri-
chers, and other factors, which are all subject to change.
Despite their levels dropping off, Fig. 8.3 shows that secondary supplies are pro-
jected to contribute 200,000 tU (over 500 million pounds U3O8) toward meeting
reactor requirements over the 2015�2030 period. Further, these supplies will con-
tinue to meet a portion of reactor needs well beyond 2030. It is likely that under-
feeding, tails enrichment, and recycle of MOX and ERU fuel will continue, at least
to some degree, with Russia continuing to be the primary source of tails enrichment
and underfeeding. Developments in enrichment and reactor technology will affect
the availability and usability of secondary supplies to some extent, but any advance-
ment is more likely to impact secondary supplies in the post-2030 period.
aConsidered a conventional resource in Brazil and is thus included in conventional resource figures.bIncludes an unknown quantity of uranium contained in monazite.Source: OECD 2014 Red Book.
240 Uranium for Nuclear Power
and diammonium phosphate in a single integrated process. The mine was
expected to produce 970 tU/year from 2015, and ramp up to 1270 tU/year in 2017
as byproduct or coproduct of phosphate. Reserves are 76,000 tU at 0.08% U, though
resources are reported as 140,000 tU at Santa Quiteria and 80,000 tU at Itataia,
grading 0.054% U in P2O5.
In the United States, Cameco and Uranium Equities Ltd have run a demonstration
plant using a refined process—PhosEnergy—and estimate that some 7700 tU
could be recovered annually as byproduct from phosphate production, including
2300 tU in the US. The prefeasibility study on the PhosEnergy process was
completed early in 2015 and confirmed its potential as a low-cost process.
9.3 Processing phosphates
Phosphate rock (phosphorite) is a marine sedimentary rock that contains 18�40%
P2O5, as well as some uranium with all its decay products, often 70�200 ppm U,
and sometimes up to 800 ppm. The main mineral in the phosphate rock is
apatite, and most commonly, fluorapatite—Ca5(PO4)3F or Ca10(PO4)6(F,OH)2. This
is insoluble, so cannot directly be used as a fertilizer (unless in very acid soils) so it
must first be processed. This is normally completed in a wet process phosphoric
acid (WPA) plant, where it is first dissolved in sulfuric acid. About 2�4% fluorine
is usually present. There are about 400 plants using this wet process worldwide,
with capacity of some 50 Mt P2O5 per year.
Some phosphate deposits—about 4% of total known—are igneous, created
by magmatic extrusion as an alkaline chimney. The main mineral is apatite, with
some fluorapatite.
When phosphate rock is treated with sulfuric acid in substoichiometric
quantity, normal superphosphate is formed. If more sulfuric acid is added,
a mixture of phosphoric acid and gypsum (calcium sulfate) is obtained. After the
gypsum is filtered out, the resultant phosphoric acid can be treated to recover
This is because most significant thorium deposits occur in deposits containing other
mineral resources of value, such as rare earth elements (REEs) and (or) titanium, as
examples.● The spent fuel waste products of thorium fission are not the types used in nuclear weap-
onry (such as plutonium, a byproduct of uranium power generation).● Thorium-rich spent fuels contain fewer radioactive elements and are smaller in volume
and mass than conventional uranium-based nuclear wastes.● The thorium fuel cycle provides an efficient way to reduce existing plutonium stocks by
using plutonium to initiate the thorium fission chain reaction.
This chapter: (1) summarizes past and present research on thorium as a nuclear
fuel that could generate electricity from small and (or) large reactors; (2) describes
the geochemistry, mineralogy, and geology of thorium deposits; (3) explains the
potential for byproduct recovery of thorium from deposits of other valuable mineral
resources; and (4) describes current research and future goals for utilizing thorium
as a nuclear fuel source.
10.2 Thorium fuel cycle
Even though thorium has been considered as a sustainable fuel cycle option, due to
the abundance of uranium and its relative ease of handling, serious attention was
not paid in the past toward developing a commercial thorium fuel cycle. Recently,
focus has renewed toward thorium utilization because of the favorable aspects of
thorium fuel explained previously.
Some countries, such as China and India, have ambitious programs in nuclear
power, but have limited resources of uranium and considerably larger resources of
thorium. In terms of energy security, utilization of thorium is favored in these coun-
tries (Anantharaman et al., 2008). Other countries, including Norway and the
United States, seek to maximize energy potential by utilization of abundantly avail-
able, domestic thorium resources.
The thorium fuel cycle differs from the uranium fuel cycle in a number of
ways. Natural thorium contains only trace amounts of fissile material (such as231Th), which are insufficient to initiate a nuclear chain reaction. In a Th-fueled
reactor, 232Th absorbs neutrons to eventually produce 233U. The 233U either fis-
sions in situ or is chemically separated from the used nuclear fuel and formed into
new nuclear fuel. The sustained fission chain reaction could be started with exist-
ing 233U or some other fissile material such as 235U or 239Pu. Subsequently, a
breeding cycle similar to, but more efficient than, that with 238U�239Pu can be
created (IAEA, 2005; WNA, 2015a).
10.2.1 Advantages of the thorium fuel cycle
Thorium-based fuels exhibit several attractive nuclear properties relative to
uranium-based fuels, such as:
● In a thermal reactor, fertile conversion of thorium is more efficient than that of 238U.● Fewer nonfissile neutron absorptions take place and neutron economy is improved.
254 Uranium for Nuclear Power
● Th fuels can be the basis for a thermal breeder reactor.● Thorium dioxide (ThO2) has a higher melting point, higher thermal conductivity, lower
coefficient of thermal expansion, and greater chemical stability than U-Pu fuel.
In the thermal spectrum, the reproduction factor (η) (number of neutrons pro-
duced per neutron absorbed in the fissile nuclide) is larger for 233U than for either235U or 239Pu. Of these, 239Pu has the lowest η in the thermal spectrum (US
Atomic Energy Commission, 1969). Although thermal breeders have no limitation
in the possibility for using any available fissile fuel, the use of 233U may be pref-
erable because of its superior neutronic characteristics (IAEA, 1979). The possibil-
ity to breed fissile material in slow neutron systems is a unique feature for
thorium-based fuels and is not possible with uranium fuels. The relatively large η
for 233U in the thermal spectrum is the most important factor contributing to the
potentially larger conversion ratios achievable with the thorium cycle. But for
realizing thermal breeding with thorium, the neutron economy in the reactor has
to be very good.
Nuclear reactor designs can be of two basic types, employing either homoge-
nous or heterogeneous core configurations. In a homogenous reactor core, nuclear
properties such as neutron flux and average cross sections are spatially uniform,
whereas nuclear properties change from one region to another in the heteroge-
neous cores (Jevremovic, 2009). In a homogenous reactor, the active core consists
of a homogenous mixture of nuclear fuel and a moderator (Lane et al., 1958). On
the other hand, in a heterogeneous reactor, there is a sharp boundary between
nuclear fuel, the moderator, and the coolant. The thorium cycle tends to be more
economical than the uranium cycle in high-temperature reactors in which the fuels
are homogenous to neutrons. In relatively homogenous thermal spectrum reactors,
such as the HTGR and the MSR, which operate at higher temperatures, the repro-
duction factor (η) remains high and the conversion ratio larger with the thorium
cycle than when using the uranium cycle. Thus, the thorium cycle tends to lower
the fuel cycle costs, even though the fuel manufacturing costs are relatively on
the higher side.
The uranium cycle tends to be economical in soft-spectrum reactors heteroge-
neous to neutrons, such as the light water reactor (LWR) and heavy water reactor
(HWR), in which the fissile enrichment requirement is low. 238U has larger reso-
nance absorption (absorption by epi-thermal neutron) than 232Th, and to reduce the
absorption by 238U, heterogeneous core configuration is required when the fuel
region is small. On the other hand, in homogenous configuration, resonance absorp-
tion occurs over a wide fuel region.
The problem of uneven power distributions can be less severe in heterogeneous
reactors utilizing the thorium cycle. The higher conversion ratio of this cycle leads
to smaller changes in power due to depletion effects. In addition, the thermal cross
section characteristics of 233U are closer to those of 235U than are those of 239Pu
to 235U. When 239Pu and 235U are present together, the 239Pu will burn out propor-
tionately faster because of its relatively larger cross section (Van Den Durpel
et al., 2011).
255Thorium as a nuclear fuel
The thorium offers an advantage from the waste perspective. Because the mass
number of 232Th is 6 units less than that of 238U, the production of plutonium (Pu)
and minor actinides (MA) (neptunium, americium, curium), which are the major
contributors to the radiotoxicity of the wastes in the uranium-plutonium cycle, is
drastically reduced if actinides are recycled (David et al., 2007). Long-term radio-
toxicity of spent fuel is mainly from Pu and MA, which is significantly lower in
Th-233U fuel cycle; this effect becomes observable only when the fuel involves Pu
or 238U. Elimination of at least the transuranic portion of the nuclear waste problem
is possible in the thorium fuel cycle, but there are some actinides that constitute a
somewhat “long-term” radiological impact, such as 229Th (half-life 7340 years).229Th is a daughter of 233U (half-life 159,200 years), which can be easily eliminated
if actinide recycle is the option.
In uranium fuels, decay (residual) heat is due to transuranium radionuclides after
100 years, in particular the Pu isotopes. In both Th-Pu and Th-U fuels, the influence
of fission products in decay heat is dominant for 100 years, and after that, actinides
are mainly responsible for the decay heat (IAEA, 2005). Decay heat is similar for
both U and Th fuels, if they are not recycled.
The 233U produced in thorium fuels is inevitably contaminated with 232U, a hard
gamma emitter; therefore, heavily shielded facilities are required for handling it.
Due to this issue, thorium-based used nuclear fuels possess inherent proliferation
resistance. However, the chemical separation of 233Pa can be possible, which can
avoid the 232U problem (Ashley et al., 2012). But chemical separation of 233Pa
from Th has not been established as yet, and even if reprocessing of highly radioac-
tive fuel is possible within 6 months, 233Pa decays with half-life of 27 days, and
decreases to less than 1%. Therefore, acquiring enough 233Pa would, in practice, be
a difficult proposition. Thorium blankets irradiated in FBR can potentially have
very low levels of 232U, if the blankets are located away from the core
(Keshavamurthy and Mohanakrishnan, 2011). This also has not been yet demon-
strated on a large scale.
10.2.2 Disadvantages of the thorium fuel cycle
Some of the unique features of the thorium fuel cycle often prove to be the major
challenges in its application. Initial fissile requirements for 235U, 233U, or Pu make
thorium unsuitable for rapid expansion of nuclear energy. Thorium introduction
could be preferred only after a good stock of fissile material (in form of either Pu
or 233U) has been built up (IAEA, 2012).232Th breeds into 233U over a comparatively long interval because 233Pa, the
intermediate isotope, has a half-life of about 27 days. As a result, some amount of233Pa builds in thorium-based fuels. 233Pa is a significant neutron absorber, and
although it eventually breeds into fissile 235U, this requires two more neutron
absorptions, which degrades neutron economy and increases the likelihood of trans-
uranic production. Hence, it is often difficult to design thorium fuels that produce
more 233U in thermal reactors than the fissile material they consume, even though
the use of 233U in a thermal reactor makes it possible to achieve higher fuel
256 Uranium for Nuclear Power
conversion ratios and longer fuel burnups than is practical with either 235U or239Pu. But the effect of 233Pa is not a major problem in some reactor designs, as it
quantitatively estimated to be only 1/1000 by weight in Th fuel. For example, in
MSR, it was demonstrated that the Breeding Ratio (BR) was 1.07 with 233Pa
removal by online reprocessing, which was reduced only to 1.05 without the 233Pa
removal (Briggs, 1967).
If thorium is used in an open fuel cycle (ie, utilizing 233U in situ), higher burnup
is necessary to achieve a favorable neutron economy. Although ThO2 has per-
formed well at burnups of 170,000 MWd/t and 150,000 MWd/t, there are challenges
associated with achieving this burnup in LWRs. These high burnup levels are
beyond those currently attainable in LWRs with zirconium cladding. High concen-
tration of 235U or Pu is required to increase burnup. While 235U enrichment of more
than 19.9% is not desirable from a proliferation perspective, high Pu is not desirable
from a waste point of view, as this negates the advantages of having lesser concen-
trations of MA, the main contributor for long-term radiotoxicity in spent fuel. In
general, the merit of incorporating thorium in open fuel cycle is often questioned
(eg, see Ashley et al., 2014).
If thorium is used in a closed fuel cycle in which 233U is recycled, remote han-
dling is necessary because of the high radiation dose resulting from the decay pro-
ducts of 232U. This is also true of recycled thorium because of the presence of228Th, which is generated from decay of 232U, but is only 1% of 232U. Although
there is substantial worldwide experience recycling uranium fuels (eg, Plutonium
and Uranium Recovery by EXtraction (PUREX)), similar technology for thorium
(eg, Thorium Recovery by EXtraction (THOREX)) is still immature. Remote
handling for the fuel fabrication will also make the fuel cycle more expensive.
10.2.3 Fast reactor thorium utilization
Fast fission of 232Th is possible, but it is only a 10th as efficient as 238U. Fissile
plutonium produced from 238U offers the potential of high breeding gains in fast
reactors with the production of 40�50% more fissile fuel than is consumed. BRs of
1.009�1.115 can be obtained with 233U in a very high-energy, fast neutron spec-
trum reactor, but it is less than that of 238U (1.284�1.582). On the other hand, in a
degraded (10�100 keV) fast spectrum, 233U would probably be as good as, or better
than, 239Pu.
The variation of 233U and 239Pu cross sections with energy is such that improved
reactivity coefficients would be obtained with the use of 233U in a sodium-cooled
FBR. This leads to improved nuclear safety characteristics. However, a fast reactor
using Th/233U cycle is not considered to have a significant advantage over thermal
reactors, even if the safety aspects are better. To breed 233U efficiently, a Pu-fueled
fast reactor with thorium as blanket would be preferable.
While thorium appears to have the nuclear disadvantages in fast spectrum, its
physical behavior as a metallic fuel material is far better than that of uranium.
Consequently, the use of thorium as a fertile material in fast spectrum reactors may
offer some advantages if it is used in the metallic form.
257Thorium as a nuclear fuel
10.2.4 Accelerator-driven systems
A subcritical reactor is a nuclear fission reactor that produces fission without
achieving criticality. Instead of a sustaining chain reaction, a subcritical reactor
uses additional neutrons from an outside source. The neutron source can be a parti-
cle accelerator producing neutrons by spallation. Such a device with a reactor cou-
pled to an accelerator is called an accelerator-driven system (ADS). Different fuel
cycles that include thorium have been proposed with these systems. The concept of
using an ADS that is based on the thorium-233U fuel cycle has been proposed by
several studies (Carminati et al., 1993; Rubbia, 1995; Furukawa, 1997).
In an ADS, the long-lived transuranic elements (such as Np and Am) in nuclear
waste can in principle fission, releasing energy in the process and leaving behind
the fission products that are shorter lived. This would shorten considerably the time
for disposal of radioactive waste (IAEA, 2003).
Various ADS designs propose a high-intensity proton accelerator with energy of
about 1 GeV, directed toward a spallation target made of thorium (either thorium
containing molten salt or solid thorium target that is cooled by liquid lead-bismuth
in the core of the reactor). In that way, for each proton interacting in the target,
20�40 neutrons on average are created to irradiate the surrounding fuel. Thus, the
neutron balance can be regulated so that the reactor would be below criticality if
the additional neutrons by the accelerator were not provided. Whenever the neutron
source is turned off, the reaction ceases.
There are technical difficulties to overcome before ADS can become economical
and eventually integrated into nuclear energy systems. The accelerator must provide
a high-intensity proton beam and should be highly reliable. There are concerns
about the accelerator-reactor coupling, especially the beam window and the spall-
ation target, which are subject to complex degradation phenomena. Beam window
is an integral part of the target containment structure. Due to combined thermos-
mechanical loads, high-energy particle irradiation, and contact with liquid heavy
metals, these components are subjected to complex stress, corrosion, and irradiation
conditions and may require frequent replacement. An emergency situation of losing
containment can be caused by window rupture.
10.3 Previous work on the thorium fuel cycle
Research toward utilizing thorium as a nuclear fuel has occurred for over 50 years.
Basic research and development, as well as operation of test reactors with thorium
fuel, has been conducted in Canada, Germany, India, Japan, the Netherlands,
Norway, the Russian Federation, Sweden, Switzerland, the United Kingdom, and
the United States (also see Martin, 2012, for a detailed history of past thorium-
reactor research in the United States). Historical thorium utilization in various reac-
tors is shown in Table 10.1 and discussed next.
258 Uranium for Nuclear Power
Table 10.1 Thorium utilization in experimental and nuclear power reactors
No Name and country Type Power Fuel Operation period
1 Indian Point 1, United States LWBR PWR, (Pin
Assemblies)
285 MWe Th1 233U Driver Fuel, Oxide
Pellets
1962�80
2 Shippingport, United States LWBR PWR, (Pin
Assemblies)
100 MWe Th1 233U Driver Fuel, Oxide
Pellets
1977�82
3 Elk River, United States BWR 24 MWe Th1 235U-238U Fuel, Oxide
Pellets
1963�68
4 Lingen, Germany BWR Irradiation-testing 60 MWe Test Fuel (Th,Pu)O2 pellets Terminated in 1973
5 KAPS 1 & 2; KGS 1 & 2;
RAPS 2, 3 & 4, India
PHWR, (Pin
Assemblies)
220 MWe ThO2 Pellets (For neutron flux
flattening of initial core after
start-up)
Continuing in all
new PHWRs
6 Peach Bottom, United States HTGR, Experimental
(Prismatic Block)
40 MWe Th1 235U Driver Fuel, Coated
fuel particles, Oxide &
dicarbides
1966�72
7 Fort St Vrain, United States HTGR, Power
(Prismatic Block)
330 MWe Th1 235U Driver Fuel, Coated
fuel particles, Dicarbide
1976�89
8 AVR, Germany HTGR, Experimental
(Pebble-bed reactor)
15 MWe Th1 235U Driver Fuel, Coated
fuel particles, Oxide &
dicarbides
1967�88
9 THTR-300, Germany HTGR, Power
(Pebble Type)
300 MWe Th1 235U, Driver Fuel, Coated
fuel particles, Oxide &
dicarbides
1985�89
10 Dragon, UK OECD-Euratom
also Sweden, Norway, and
Switzerland
HTGR, Experimental
(Pin-in-Block Design)
20 MWt Th1 235U Driver Fuel, Coated
fuel particles, Oxide &
Dicarbides
1966�73
(Continued)
Table 10.1 (Continued)
No Name and country Type Power Fuel Operation period
11 MSRE ORNL, United States MSR 7.5 MWt 233U Molten Fluorides 1964�69
12 FBTR, India LMFBR, (Pin
Assemblies)
40 MWt ThO2 blanket 1985�present
13 SUSPOP/KSTR KEMA,
Netherlands
Aqueous Homogenous
Suspension (Pin
Assemblies)
1 MWt Th1HEU, Oxide Pellets 1974�77
14 NRU & NRX, Canada MTR (Pin Assemblies) � Th1 235U, Test Fuel Irradiation-testing of
a few fuel
elements
15 KAMINI, India MTR Thermal 30 kWt Al1 233U Driver Fuel, Research reactor in
operation
1996�present
16 CIRUS, India MTR Thermal 40 MWt “J” rod of Th & ThO2 Research reactor
1960�2010
17 DHRUVA, India MTR Thermal 100 MWt “J” rod of ThO2 Research reactor in
operation
1985�present
MWe, Megawatt electric. Reactor types: LWBR, light water breeder reactor; PWR, pressurized water reactor; BWR, boiling water reactor; PHWR, pressurized heavy water reactor;
HTGR, high-temperature gas-cooled reactors; MSR, molten salt breeder reactor; LMFBR, liquid metal fast breeder reactor; MTR, materials testing reactor.
10.3.1 Light water reactors
LWRs can be operated with thorium fuel. In the United States, thorium fuel was
tested in LWBR pressurized water reactors (PWRs) at the Indian Point plant in
New York initially, but it was later converted to uranium fuel cycle. Shippingport
reactor in Pennsylvania, a “seed-blanket” LWBR PWR, operated with thorium fuel
between August 1977 and October 1982. At the end of core life, 1.39% more fissile
material was present, proving that LWR breeding was possible with thorium.
Boiling water reactors (BWR) offer a design flexibility that can be optimized for
thorium fuels. Thorium was also used in the BWR at Elk River Reactor in Minnesota,
United States. Thorium fuel was also tested at the 60 MWe BWR in Lingen,
Germany.
Germany and Brazil collaborated on various theoretical and some experimental
studies in thorium utilization in PWRs during 1979�88; they concluded that for
once-through operation the U/Th cycle was as competitive as the U/U cycle
(Pinheiro et al., 1989). Studies by EDF in France also arrived at similar conclusions
(IAEA, 1985). Studies were also carried out in Russia for thorium utilization in
water�water energetic reactors (Morozov et al., 1999).
10.3.2 Heavy water reactors
HWRs could offer excellent neutron economy and faster neutron energy, so they
are considered better for breeding 233U. Conceptual design studies have indicated
that thorium and uranium fuel concepts have many common design characteristics
and that the thorium cycle could be used in a plant designed for the uranium cycle
without large performance penalties. In Canada, Atomic Energy Canada Limited
has more than 50 years of experience with thorium-based fuels, including burnup to
47 GWd/t. So far, about 25 tests have been performed in three research reactors and
one precommercial reactor. India is continuing the use of ThO2 pellets in PHWRs,
for neutron flux flattening of initial core after start-up.
10.3.3 Fast breeder reactors
There is no relative advantage in using thorium instead of depleted uranium as a
fertile fuel matrix in FBR systems due to a higher fast fission rate for 238U and the
fission contribution from residual 235U in this material. The fast fission cross sec-
tion of 238U is significantly larger than that of 232Th by a factor of 4�5. Because
large stocks of depleted uranium are available as tails in uranium enrichment, there
is little incentive to develop the use of thorium in these reactors.232Th in the blanket can be advantageous in a mixed reactor scenario. India has
a three-stage nuclear energy scenario in which FBR play an important role; the
use of thorium in the blanket to breed 233U was tested in a 40 MWt fast breeder
test reactor (FBTR), Kalpakkam, which has been operational since 1985. The
Kamini 30 kWt experimental neutron-source research reactor, adjacent to the
FBTR, uses 233U as fuel. Thorium blanket is planned to be used in a 500 MWe
261Thorium as a nuclear fuel
prototype fast breeder reactor (PFBR), which is expected to be operational by
2015 (WNA, 2015b).
While the use of the 232Th- 233U cycle in a FBR does not, in general, appear to
be as attractive as the 238U- 239Pu cycle, the use of 233U in the core may provide
advantages in reactor safety and control. The use of either thorium or 233U in a fast
spectrum can mitigate the issue of positive sodium void coefficient that is obtained
with the uranium cycle under corresponding design conditions.
Studies were done in Russia on thorium utilization in BN-type fast reactors.
Utilization of Pu/Th fuel has been proposed for the BN-800 type reactor (expected
to be commercially operational in 2015).
10.3.4 High-temperature gas-cooled reactors
HTGRs are thermal spectrum reactors moderated with graphite and cooled by
helium. Tristructural-isotropic (TRISO) fuel, composed of particles of thorium
mixed with plutonium or enriched uranium, is the fuel in pebble-bed or prismatic
arrangement. Deep burn of the fissile content and flexibility in fuel management
makes high-temperature gas-cooled reactors (HTGR) a good option for thorium
utilization. The possible use of beryllium oxide (BeO) in the fuel element, either as
a matrix for the fuel particles or in place of some of the bulk graphite in the
element, would greatly enhance the conversion ratio of the system.
In Germany, a 15 MWe AVR (Arbeitsgemeinschaft Versuchsreaktor) experimental
pebble-bed nuclear reactor at Julich operated from 1967 to 1988, partly as a test bed
for various fuel pebbles, including thorium. The 300 MWe thorium high-temperature
reactor (THTR), developed from the AVR, operated between 1983 and 1989 with
674,000 pebbles, over half containing Th/highly enriched uranium (HEU) fuel.
In the United States, General Atomics’ Peach Bottom high-temperature,
graphite-moderated, helium-cooled reactor operated between 1967 and 1974 at
110 MWth, using high-enriched uranium with thorium. The Fort St Vrain reactor,
the only commercial thorium-fueled nuclear plant in the United States, was a high-
temperature (700�C), graphite-moderated, helium-cooled reactor with a Th/HEU
fuel designed to operate at 842 MWth (330 MWe). The fuel in the Fort St Vrain
reactor was arranged in hexagonal columns (“prisms”) rather than as pebbles.
Almost 25 tTh were used as fuel for the reactor, and this achieved 170 GWd/t burn-
up; the reactor operated from 1979 to 1989.
In the United Kingdom, thorium fuel elements with a 10:1 Th/U (HEU) ratio
were irradiated in the 20 MWth Dragon reactor at Winfrith for 741 full power days.
The Dragon reactor ran between 1964 and 1973 as an OECD/Euratom cooperation
project, involving Austria, Denmark, Sweden, Norway, and Switzerland, in addition
to the United Kingdom.
Because there was no established thorium ceramic fuel reprocessing route, com-
mercial attention to the THTR has been directed to a once-through mode of opera-
tion. Even though this requires nearly double the amount of 235U over a reactor
lifetime compared with the 233U recycle case, the 235U requirement is much less
than that of the current LWRs operating in the once-through mode.
262 Uranium for Nuclear Power
10.3.5 Molten salt reactors
MSRs offer attractive concepts for thorium utilization. The primary coolant, and
even the fuel itself, is a molten salt mixture, with graphite usually used as the mod-
erator (Serp et al., 2014). In theory, this design can operate as a full recycle system,
with continuous “online” pyroprocessing for separation of fission products. Some
designs of MSRs are capable to “burn” problematic MA. MSRs may operate with
epithermal or fast neutron spectrums, and with a variety of fuels, but good neutron
economy makes the MSR attractive for the thorium fuel cycle.
The molten salts coolant, which is mostly lithium-beryllium fluoride (called
FLiBe, 2LiF-BeF2), remain liquid without pressurization from about 500�C up to
about 1400�C, in marked contrast to a PWR which operates at about 315�C under
150 atmospheres pressure. The advanced MSR concept is to have the fuel dissolved
in the coolant as fuel salt. Intermediate designs have fuel particles in solid graphite
and have less potential for thorium use (WNA, 2015c). Safety is achieved with a
freeze plug, which if power is cut, allows the fuel to drain into a drain tank, which
has a subcritical geometry. There is also a negative temperature coefficient of reac-
tivity due to expansion of the fuel.
In 1954, Oak Ridge National Laboratory (ORNL), in the United States, con-
ducted an aircraft reactor experiment (ARE), which was a 2.5 MWt nuclear reactor
experiment designed to attain a high power density for use as an engine in a
nuclear-powered aircraft (Rosenthal, 2009). The ARE used molten fluoride salt
NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, moderated by BeO, with helium gas as a
secondary coolant; the reaction had a peak temperature of 880�C. The experiments
ran for only several days. The Pratt and Whitney Aircraft Reactor No. 1 (PWAR-1)
was a zero power MSR that was tested in ORNL in 1957, which used NaF-ZrF4-
UF4 as the primary fuel and coolant (Scott et al., 1958).
ORNL continued this work with the molten salt reactor experiment (MSRE), a
7.4 MWt test reactor (Briggs, 1964). It went critical in 1965 and ran for 4 years.
The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-29-5-1), the graphite core mod-
erated it, and its secondary coolant was fertile blanket of lithium-beryllium fluoride
(FLiBe) (2LiF-BeF2). It primarily used two uranium fuels: first 235U and later 233U.
The latter 233UF4 was the result of breeding from thorium in other reactors.
Because this was an engineering test, the large, expensive breeding blanket of tho-
rium salt was omitted in favor of neutron measurements.
During 1970�76, ORNL proposed a design for a 1000 MWe molten salt breeder
reactor (MSBR) that would use LiF-BeF2-ThF4-UF4 (72-16-12-0.4) as fuel and be
moderated by graphite (MacPherson, 1985). The MSBR offered the potential of a
BR of 1.07, a specific inventory on the order of 1.0 kg fissile/MWe or less, a power
doubling time of less than 15 years. Even though the BR was smaller than that of
the fast breeder designs proposed at that time, the doubling time was almost the
same (Yoshioka, 2013). However, this program in the United States was closed in
favor of the liquid metal fast breeder reactor.
The UK conducted theoretical and experimental research on a lead-cooled,
2.5 GWe molten salt fast reactor (MSFR) using chloride salt and fueled by
263Thorium as a nuclear fuel
plutonium, and either helium gas or molten lead as the secondary coolant (Endicott,
2014). Theoretical work on the concept was conducted between 1964 and 1966,
and experimental work was done between 1968 and 1973. Russia conducted some
theoretical and experimental research on MSRs during the 1970s at the Kurchatov
Institute (Novikov, 1995).
As molten salts are present as eutectic mixture of LiF-BeF2, with fertile Th and
fissile U/Pu dissolved in the medium, it serves as fuel element, heat transfer, and a
fuel-processing medium. For reprocessing the solid spent fuel, PUREX is commer-
cially available. In this process, spent fuel elements are dissolved as aqueous solution
from which Pu and U are recovered by organic solvent extraction. However, molten
salt processing can be possible through the “dry” pyroprocessing route developed in
recent years, which could be simpler and compact than the PUREX process (Yoshioka,
2013). Dry processing technology was proposed and developed at Oak Ridge National
Laboratory (ORNL) in Oak Ridge, Tennessee (US) in the 1970s and later in European
countries; this technology was tried for the solid fuels. The online processing of molten
fuel salts in MSR designs usually involves removal of U, Pa and FP (Fission Products)
and reintroduction of U, Pu, or Th. MA remain in the reactor until they fission.
10.3.6 Resource requirements
Because 232Th is not fissile, a certain quantity of 235U, 233U, or Pu is required in
the fuel cycle. The amount of fissile material required depends on the reactor sys-
tem and the fuel cycle options (once-through, limited recycle, or continuous recy-
cle). Natural uranium as a starting point for 235U and Pu will be required for
thorium-based nuclear energy systems. Depending on the system, natural uranium
will be required either in the driver fuel or as a continuous top-up.
A thorium fuel cycle can be self-sustaining if enough 233U is created to replace
the quantities that are consumed. It will remain necessary to provide fissile material
from elsewhere (235U or plutonium) to provide inventories for new construction. To
achieve this self-sustaining cycle, fuel burnup has to be limited to about
13,000 MWd/t HM in the case of a PHWR. This is called the self-sustaining equi-
librium thorium cycle. Depending on the various costs involved, however, it will be
more economical to aim for higher fuel burnup and accept the need for a continuing
supply of fissile material “driver” for topping up.
Taking a “high burnup” case of 40,000 MWd/t HM and assuming that its pluto-
nium requirements are to be supplied by natural uranium CANDUs, the thorium-
using reactors would use only about 30 tTh, plus 140 Kg Pu/year (generated by 45 t
of natural uranium), in comparison to 140 t of natural uranium per GW-year for a
once-through natural uranium cycle. A mixed fuel cycle of this type could be com-
petitive in resource utilization (Lung, 1997).
Wigeland et al. (2014) evaluated various options in nuclear fuel cycles in terms
of different criteria, including fuel resources utilization. Of the 40 evaluation groups
considered in the study, 15 have thorium as a fertile material in the fuel cycle.
Natural thorium and uranium resources requirements for the various evaluation
groups in the study are summarized in Table 10.2.
264 Uranium for Nuclear Power
Table 10.2 Natural thorium and uranium resources requirement for various utilization scenarios
Evaluation
group
Short description System considered Natural
Th use
(t/GWe)
Natural
U use
(t/GWe)
Once-through
EG05 Once-through using enriched-U/Th fuel in
thermal or fast critical reactors
High-Conversion HTGR with LEU and Th fuel 4.65 289.20
EG06 Once-through using Th fuel to very high burnup
in thermal EDS
Breed and Burn 233U/Th in Thermal spectrum
FFH
9.88 0.00
EG08 Once-through using Th fuel to very high burnup
in fast EDS
Breed and Burn 233U/Th in Fast spectrum
Fusion-Fission Hybrid (FFH)
1.62 0.00
Limited recycle
EG10 Limited recycle of 233U/Th with new Th fuel in
fast and/or thermal critical reactors
Limited recycle of 233U/Th in MSR 10.86 0.00
EG11 Limited recycle of 233U/Th with new enriched-
U/Th fuel in fast or thermal critical reactors
Breed and Burn 233U/Th with LEU Support in
sodium-cooled fast reactor (SFR) with Partial
Separation
2.55 106.80
EG17 Limited recycle of Pu/Th with new enriched-U/
Th fuel in thermal critical reactors
Limited recycle of Pu from PWR in an PWR
burner fueled with Thorium
1.88 172.41
EG18 Limited recycle of 233U/Th with new enriched-
U/Th fuel in thermal critical reactors
Limited recycle of 233U/Th from PWR in a
PWR burner
3.42 152.16
Continuous recycle
EG25 Continuous recycle of 233U/Th with new
enriched-U/Th fuel in thermal critical reactors
Continuous recycle of 233U/Th in PWR with
LEU Support
0.85 113.54
(Continued)
Table 10.2 (Continued)
Evaluation
group
Short description System considered Natural
Th use
(t/GWe)
Natural
U use
(t/GWe)
EG26 Continuous recycle of 233U/Th with new Th fuel
in thermal critical reactors
Continuous recycle of 233U/Th in MSR 1.25 0.00
EG27 Continuous recycle of 233U/Th with new
enriched-U/Th fuel in fast critical reactors
Continuous recycle of 233U/Th in SFR with
LEU Support
0.45 186.62
EG28 Continuous recycle of 233U/Th with new Th fuel
in fast critical reactors
Continuous recycle of 233U/Th in SFR 1.68 0.00
EG37 Continuous recycle of 233U/Th with new
enriched-U/Th fuel in both fast and thermal
critical reactors
Continuous Recycle of TRU/U from PWR in
SFR and produce 233U/Th for recycle in
Advanced PWR
0.43 24.36
EG38 Continuous recycle of 233U/Th with new Th fuel
in both fast and thermal critical reactors
Continuous Recycle of 233U/Th produced in
SFR in PWR
1.93 0.00
EG39 Continuous recycle of 233U/Th with new
enriched-U fuel in both thermal critical
reactors and fast EDS
Continuous Recycle of 233U in PWRs and burn
TRU in ADS
0.75 114.85
EG40 Continuous recycle of 233U/Th with new Th fuel
in fast EDS and thermal critical reactors
Produce 233U/Th in ADS and continuously
recycle in PWR
1.51 0.00
The maximum natural thorium requirement of 11 t/GWe is estimated in order for
a limited recycle of 233U/Th in a MSR system. Other scenarios with limited recycle
indicate thorium requirements of 1.88�3.42 t/GWe, but all scenarios have higher
natural uranium use of 107�172 t/GWe. Some of the continuous recycle scenarios
also have a very low thorium requirement, as well as no or limited natural uranium
requirement, such as the continuous recycle in a MSR.
10.4 Current research and future possibilities
Despite many projects and pilot test reactors in the second half of the 20th century, the
use of thorium as a reactor fuel has yet to be commercialized in a modern power reac-
tor. However, today ongoing research and development on advanced reactor designs
may employ thorium as a nuclear fuel, described next. These projects include HTGR;
MSR; CANDU-type reactors; AHWR; FBR, and PHWR.
10.4.1 China
China in collaboration with the USA has very active ongoing research on thorium
utilization in MSR designs (Li et al., 2015). This is a dual program involving an
early solid fuel stream and advanced liquid fuel stream (WNA, 2015c). In January
2011, the China Academy of Sciences launched a research and development pro-
gram on a liquid-fluoride thorium reactor (LFTR), called the thorium-breeding mol-
ten salt reactor (TMSR). In 2013, the National Energy Administration included the
TMSR project among the 25 “National Energy Major Application-Technology
Research and Demonstration Projects” in its “Plan of Energy Development
Strategy.” In 2014, the local government of Shanghai launched a major TMSR proj-
ect to support the TMSR technology development.
The TMSR program is divided into three stages. In the early stage, a 10 MWt
solid-fueled molten salt test reactor (TMSR- SF1) and a 2 MWt liquid-fueled mol-
ten salt experimental reactor (TMSR-LF1) are planned for construction and opera-
tion by 2016. In the engineering experimental stage, a 100 MWt solid-fueled TMSR
demonstration system (TMSR-SF2) is planned by 2025 and a 10 MWt liquid-fueled
molten salt experimental reactor (TMSR-LF2) is planned by 2018. The third indus-
trial promotion stage will aim for the commercialization of a 1 GW TMSR-SF3 by
2030. A fast spectrum TMSFR-LF fast reactor optimized for burning of MA is also
envisaged.
Solid fuel MSR technology was preferred in the early stage, due to the technical
difficulty associated with high radioactivity of the molten salt when they contain
dissolved fuels and wastes. After the accumulation of experience is gained with
component design, operation, and maintenance of clean salts, use of liquid salt will
be applied. Molten salt fuel is considered superior to the TRISO fuel in effectively
unlimited burnup, less waste, and lower fabricating cost (WNA, 2015c).
267Thorium as a nuclear fuel
Solid fuel envisages only partial utilization of thorium with an open fuel cycle,
whereas liquid fuel designs will have a fully closed Th-U breeding cycle. Solid fuel
TRISO particles will be with both low-enriched uranium and thorium, separately.
The first step will be to develop solid fuel, bypassing the difficult reprocessing and
refabrication options, and subsequently mastering the complex fluid fuel technol-
ogy. The US cooperation with this project is primarily on the solid fuel technology,
which is considered as the realistic first step.
China is developing HTR-PM, which is a graphite-moderated, helium-cooled
high-temperature reactor. It is possible to use thorium in this type of reactor.
Construction of a twin HTR-PM unit started in 2014, and is expected to be opera-
tional by 2017.
10.4.2 Canada
Since 2008, CANDU Energy of Canada and the China National Nuclear Corporation
(CNNC) are cooperating in the development of thorium and recycled uranium as
alternative fuels for new CANDU reactors. CANDU Energy is now part of SNC-
Lavalin, and works on Advanced Fuel CANDU Reactor (AFCR) technology, which
aims at thorium utilization. AFCR will be designed to use recycled uranium or tho-
rium as fuel, thus reducing spent fuel inventories and significantly reducing fresh ura-
nium requirements. Spent fuel from four conventional PWR reactors can fully supply
one AFCR unit (as well as providing recycled plutonium for mixed oxide fuel
(MOX)). CNNC is currently preparing an AFCR feasibility study and proposal to
China’s National Energy Administration, to be submitted by early 2015.
The integral molten salt reactor is proposed by the Terrestrial Energy in Canada
(LeBlanc, 2013). It is based on the MSBR, but is a kind of tank-type reactor, where
heat exchangers are enclosed in the reactor vessel.
10.4.3 India
In India, research on thorium utilization has been carried out since the 1950s. A
three-stage nuclear energy program with uranium-fueled PHWRs, plutonium-fueled
FBRs, and thorium-233U-based AHWRs has been envisaged. A 500 MWe PFBR is
in the final stages of construction, and is expected to be completed in 2015. More
500 MWe FBRs are planned for immediate deployment and beyond 2025; a series
of 1000 MWe FBRs with metallic fuel, capable of high breeding potential are pro-
posed (Chetal et al., 2011). The large-scale deployment of thorium is expected to
be about three to four decades after the commercial operation of FBR with short
doubling time, when thorium can be introduced to generate 233U.
Demonstration of the use of thorium on an industrial scale is planned in an
AHWR. AHWR is a 300 MWe boiling light water cooled and heavy water moder-
ated vertical pressure tube type reactor (Sinha and Kakodkar, 2006). The reactor is
designed with the dual objective of utilization of abundant thorium resources and to
meet the future demands to nuclear power, which includes enhanced safety and reli-
ability, improved economics, and a high level of proliferation resistance. It has
268 Uranium for Nuclear Power
many passive and inherently safe features so that the reactor can be located close to
population centers.
During mid-2010, a prelicensing safety appraisal of the planned experimental
thorium-fueled 300 MW(e) AHWR was completed by the (India) Atomic Energy
Regulatory Board. The site-selection process started in 2011; the reactor is expected
to become operational by 2020. An experimental assembly with AHWR type (Th-
Pu) MOX fuel pins completed its test irradiation, and another with (Th-LEU) MOX
fuel pins has been loaded in the Dhruva research reactor. Several test facilities have
been setup for the AHWR design validation.
India is also considering the use of thorium in a compact high-temperature reac-
tor (CHTR) and innovative high-temperature reactor for hydrogen production
(IHTR). Both of the designs use 233U�Th-based TRISO coated particle fuel. CHTR
uses lead-bismuth as coolant and IHTR molten salt.
India’s nuclear energy program, in general, and thorium utilization in particular,
are heavily linked to solid fuels, its fabrication, reprocessing, and refabrication. There
are technical challenges in Th-based fuel fabrication that require higher sintering tem-
perature. Thorium-based fuels are reprocessed using the THOREX process, which is
similar to the PUREX process. The THOREX process is still in the developmental
stage and need extensive modifications prior to large-scale deployment. There are
challenges in dissolving the very inert thorium fuel in aqueous solutions. The presence
of 232U requires automated reprocessing and fuel fabrication in shielded facilities.
Breeding of 233U from 232Th goes through an intermediate 233Pa with about 27 days
half-life. Therefore, a cooling time of at least 1 year is required for maximizing the
recovery of 233U. Remaining 233Pa in the fuel will go into the waste stream, which
could have a long-term radiological impact. To avoid this, it is essential to separate233Pa from the spent fuel prior to extraction of 233U and Th (Vijayan et al., 2013).
India currently has only small uranium spent fuel reprocessing plants at three
sites with a total of 330 t/year capacity, with mostly the PUREX process in use
(except for reprocessing fast reactor carbide fuel). To close the FBR fuel cycle, a
fast reactor fuel cycle facility is under construction. Power reactor thorium repro-
cessing facility has been constructed to reprocess ThO2 fuel bundles irradiated in
PHWRs; this facility is under commissioning. India considers that premature
deployment of thorium would lead to suboptimal use of indigenous energy
resources, and that it would be necessary to build up a significant amount of fissile
material prior to launching the thorium cycle in a big way. Therefore, thorium-
based reactors are expected to be deployed only beyond 2070.
Recently, India started to also consider MSRs as one of the promising options
for thorium utilization. Conceptual design of the Indian molten salt breeder reactors
(IMSBRs) is currently under research and design. India considers IMSBRs to pro-
vide significant advantages over metallic fueled FBRs, such as reduced fissile
inventory and better BR. The major difference in practical terms will be the bypass-
ing of the comparatively difficult fuel fabrication, reprocessing, and fuel refabrica-
tion requirements.
India also has an active research program for thorium utilization in ADS systems
(Nema, 2011; Sinha, 2011).
269Thorium as a nuclear fuel
10.4.4 Norway
In April 2013, Thor Energy of Norway commenced a thorium MOX testing pro-
gram in the Halden research reactor in Norway. Fuel irradiation is being tested to
determine if thorium-plutonium (Th-Pu) MOX can be used in commercial nuclear
power plants. Thor has commenced discussions with several utilities about the use
of these thorium-mixture fuels in commercial LWRs and is conducting feasibility
studies with one utility. Thor has commenced discussions with several regulators
about the licensing of thorium fuels for use in these LWRs.
10.4.5 Europe
Safety Assessment of the Molten Salt Fast Reactor (SAMOFAR) is a consortium
that consists of 11 participants, which include universities and research laboratories,
such as CNRS (France), JRC (European Commonwealth), CIRTEN (Italy), TU
Delft (Netherlands), CINVESTAV (Mexico), and PSI (Switzerland). Industrial part-
ners include IRSN (France), AREVA (France), KIT (Germany), EDF (France), and
CEA (France). The objective of SAMOFAR is to prove the innovative safety con-
cepts of the MSFR by advanced experimental and numerical techniques.
10.4.6 South Africa
Steenkampskraal Thorium Ltd (STL), South Africa is undertaking several activities
related to the thorium fuel cycle. STL owns the rights to the thorium that will be
produced at the Steenkampskraal rare earths and thorium mine in South Africa.
STL has designed a thorium refinery for the production of reactor-grade thorium.
STL is designing a generation IV high-temperature, gas-cooled, pebble-bed reactor,
the HTMR100 (100 MWth high-temperature modular reactor). The HTMR100 will
be cooled with helium gas and operate at temperatures around 750�C.
10.4.7 United States
Several companies are developing innovative thorium fuel for LWRs, and HTGR
and MSR concepts for thorium utilization.
Lightbridge is developing advanced metallic nuclear fuel for maximum power
levels and operating cycle length extension in LWRs. The metallic fuel rod also
forms the central seed region of the seed-and-blanket thorium-based LWR fuels.
Current development efforts are focused on demonstrating the performance of the
metallic fuel in a prototypic LWR environment and developing data and methodol-
ogies that will enable evaluation of the fuel for nuclear regulatory licensing.
X-energy is designing the Xe-100 reactor, a high-temperature gas-cooled pebble-
bed nuclear reactor. Xe-100’s small footprint and safe design allow operation close
to population centers, and it is suited for installation at geographically constrained
sites. The Xe-100 is being designed specifically to accommodate a variety of fuel
types, including thorium, without design changes. X-energy’s current plans leverage
270 Uranium for Nuclear Power
the US Department of Energy’s investment in fuels through the Next Generation
Nuclear Plant program.
Transatomic Power Corp has finalized the preliminary design of an advanced
MSR, and began experimental testing of key materials and components. Corrosion,
radiation, and high-temperature materials testing are being conducted under a 3-
year sponsored research agreement with the Department of Nuclear Science and
Engineering at the Massachusetts Institute of Technology.
Martingale is developing a simple thorium MSR called ThorCon. Preliminary
detailed design has been completed.
Flibe Energy is planning to develop and commercialize a LFTR. LFTR is a het-
erogeneous MSR design that breeds its 233U fuel from a FLiBe salts with thorium
fluoride. A study of the LFTR design to understand the potential and challenges has
commenced.
10.4.8 United Kingdom
Moltex Energy’s stable salt reactor (SSR) is a conceptual design with no pumps
(only impellers in the secondary salt bath) and relies on convection from vertical
fuel tubes in the core to convey heat to the integral steam generators. The SSR can
be run with thermal or fast neutron spectrums and thorium can be potentially used.
10.4.9 Generation IV international forum
The Generation IV International Forum (GIF) has Canada, China, the European
Atomic Energy Community (Euratom), France, Japan, Russia, South Africa, South
Korea, Switzerland, and the United States as active members. The GIF explore
areas of mutual interest and make recommendations regarding research and devel-
opment areas and processes for development of generation IV nuclear energy sys-
tems. The GIF program for the MSR includes the concepts of MSFR where Th fuel
can be used. The GIF 2014 annual report (The Generation IV International Forum,
2015) said that a lot of work must be done on salts before demonstration reactors
were operational, and they suggested 2025 as the end of the viability research and
development phase.
10.4.10 Russia
Russia’s molten salt actinide recycler and transmuter is a fast reactor fueled only by
transuranic fluorides from uranium and MOX LWR used fuel. It is part of the
MARS (minor actinide recycling in molten salt) project involving Scientific
Research Institute of Atomic Reactors (Russia) (RIAR), Kurchatov, and other
research organizations. The 2400 MWt design has a homogenous core of Li-Na-Be
or Li-Be fluorides without graphite moderator and has reduced reprocessing com-
pared with the original US design. Thorium may also be used, though it is described
as a burner-converter rather than a breeder.
271Thorium as a nuclear fuel
10.4.11 Germany
The Institut fur Festkorper-Kernphysik gGmbH (Institute for Solid-State Nuclear
Physics—IFK) is designing a reactor concept called the dual fluid reactor (DFR).
The DFR can use versatile nuclear fuels, for example thorium, natural uranium, or
even depleted uranium and spent nuclear fuel. It has a hard neutron spectrum that
favors the poor neutron economy of the Th-U cycle, yielding better doubling times.
It allows for the production of custom nuclides for medical use, including the scarce
Molybdenum-99/Technetium-99m.
10.4.12 Denmark
Copenhagen Atomics plans to design a MSR-based system, preferably to fit in a
shipping container, called the Copenhagen Atomics Waste Burner (CAWB). The
CAWB will use thorium to burn out actinides from spent nuclear fuel to convert
long-lived radioactive waste into short-lived radioactive waste, while producing
energy. The aim is to optimize the CAWB so that it can start on fissile material
from spent nuclear fuel alone, or with added external fissile material. The first ver-
sion of CAWB will have a 50 MWt capacity.
10.4.13 Japan
International Thorium Molten Salt Forum (ITMSF) and Thorium Tech Solution Inc.
(TTS) are developing the FUJI MSR. The proposed design is rated at 200 MWe out-
put. The consortium plans to first build a much smaller MiniFUJI, a 7 MWe reactor of
the same design (Yoshioka, 2013). A 1000 MWe capacity Super-FUJI is also under
preliminary design phase. Continuous chemical processing of fuel salt is not per-
formed in FUJI, as the system proposed to use ADS for production of fissile 233U. But
radioactive Xe, Kr, and tritium are removed from the fuel salt continuously. The total
fissile 233U required for 30 years operation at 70% load factor is 1.48 t, and the
remaining fissile at the end of the reactor life is 1.50 t. A total of 3.9 tTh will be con-
sumed in 30 years. Estimated production of Pu will be 0.3 kg after 30 years of opera-
tion, which is 0.03% of that of a BWR with same capacity. The production of MA is
5.3 kg after 30 years, which is 5% of that of a same sized BWR.
Kyoto Neutronics is designing a small thorium MSR integrated with an accelera-
tor neutron source, which they refer to as UNOMI (Universally Operable Molten
salt reactor Integrated).
10.4.14 Thorium production for the foreseeable future
In current market conditions, primary production of thorium is not likely to be as
economic as an independent private enterprise. Recovery of thorium as a byproduct,
extracted during REE recovery from monazite [(REEs,Th)PO4)], seems to be the
most feasible source of thorium production at this time.
272 Uranium for Nuclear Power
The recovery of monazite from raw sand or crushed ore is possible by gravity
separation techniques (utilizing the high density of monazite) and electrostatic
methods (monazite displays weak magnetism). The monazite is then dissolved in
either sodium hydroxide or sulfuric acid. The resulting solutions contain REEs, ura-
nium and thorium. This is followed by a multistage process using organic phases to
achieve separation with a final product of Th oxide.
Processing of monazite to recover rare earths and thorium has been effectively
done in the past in many countries. Today, recovery of monazite (and possibly tho-
rite) as a coproduct of REE mining and processing, made efficient and relatively
inexpensive by modern heavy-mineral separation techniques, may provide an eco-
nomical means of acquiring thorium for the foreseeable future. This thorium then
may become a source of fuel stock for a new generation of nuclear power reactors.
10.5 Thorium geology and resources
10.5.1 Thorium geochemistry and mineralogy
Thorium is among a group of elements sometimes referred to as the high field
strength elements (HFSE). These are elements having a valence state greater than
two (high charge) and small-to-medium-size ionic radii, thus producing a high elec-
tric field (high field strength). Besides thorium, these elements also include the
1Retired Cameco Corporation, Port Perry, ON, Canada, 2Retired Cameco Corporation,
Saskatoon, SK, Canada
11.1 Introduction
Processing of uranium beyond the milling stage is somewhat generic. While milling
tends to be ore specific, mill concentrates are essentially fungible, and subsequent
processing is more easily transferred from one location to another. Thus the subse-
quent operations, of which refining and conversion are the first step, are subject to
much more scrutiny to prevent nuclear proliferation, as well as commercial interests
in a very competitive business. These restrictions put many constraints on the level
of technical details that can be published, including in this current chapter, which is
limited to summarizing what is appropriate for general distribution in the public
domain. Nevertheless the refining and conversion of uranium processing is
described, including some history of technology development, how this history
affected past technical and economic choices, and how these are likely to affect fur-
ther developments and activities in this area.
11.2 Conversion and needs
The extraction of uranium containing minerals occurs through leaching where the
initial purification also occurs (milling operations) at or near mine sites provides
uranium ore concentrates (UOCs or “yellowcake”). Chemically, these are most
commonly triuranium octoxide (U3O8) or uranyl peroxide (UO4), but have also
included sodium diuranate and magnesium diuranate.
The further processing of the concentrates has the two primary goals of
providing:
● Refining to achieve impurity contents in the converted uranium products, which are suffi-
ciently low that there are no serious interferences with the enrichment, fuel manufacturing,
and reactor processes, hence contributing to stability and efficiency in those operations● Conversion to a form suitable for feed to the enrichment process, on a commercial basis
that form is exclusively uranium hexafluoride, UF6: although other feed options have
been suggested for some enrichment processes, none of these are currently in commercial
operation, or even in late-stage development
The commercial purchase of the natural UF6 is made by electrical utilities, which
need a material that is deliverable to multiple enrichers, thereby encouraging com-
petition and expected better pricing for this next step in nuclear fuel production.
Conversion to UF6 is only a small part of the fuel cycle, and is often not considered
Uranium for Nuclear Power. DOI: http://dx.doi.org/10.1016/B978-0-08-100307-7.00011-9
of strategic importance. Little attention is paid to this segment yet it is clearly
essential for enrichment to take place.
Refining and conversion to natural UO2 is also commercially important although
only smaller in scale compared to conversion to natural UF6. Natural UO2 is used
for manufacture of CANDU fuel and for blanket fuel in some light water reactors.
For convenient matching with fuel manufacturing, fuel blanket material is often
made via natural UF6, but is more economically produced by a direct process
described near the end of this chapter.
11.3 Conversion technologies
11.3.1 Process outline
The refining and conversion of UOC to UF6 are discussed together as the two steps
may be in different order. The most common process involves refining and subse-
quent conversion to UF6, and this is often called the “wet process” as this refining
route requires water solution. The alternative route to UF6 involves conversion and
then refining by distillation of the UF6. Because no water is involved in the main
process steps, this is called the “dry process.”
UF6 can be generated from the reaction of virtually any uranium compound with
elemental fluorine at elevated temperature. However, fluorine is both expensive and
hazardous to store and react, so cost-effective processing minimizes its use.
Nevertheless, production of UF6 requires more commercial F2 production than any
other use of F2.
The commercial routes to UF6 minimize F2 usage by using three steps instead of
just one. These steps are reduction (reaction with H2), hydrofluorination (reaction
with hydrogen fluoride (HF)), and fluorination (reaction with F2). Thus, the steps
may be represented by the equations:
UO3 ðor other oxideÞ1H2 ! UO2 1H2O
UO2 1 4HF ! UF4 1 2H2O
UF4 1 F2 ! UF6
11.3.2 History
While downstream uranium processing today is focused on fuel supply for the
peaceful production of electrical power, the initial focus for technology develop-
ment was for nuclear weapons and a great deal of investment was made in those
early days to develop and apply processing methods as rapidly as possible. Choices
made in those early days still influence current processing routes and locations.
Even the industry operations terminology, still in use today, comes from an earlier
time when the word uranium was to be avoided for security reasons. For example,
300 Uranium for Nuclear Power
we often hear of “yellowcake” for mill concentrate (even UOC) and “green salt” as
the precursor to UF6.
The initial challenge of producing large quantities of high purity uranium in the
1940s was addressed by turning to the laboratory method of purifying uranium,
which involved the extraction of uranium from aqueous nitrate solution into ether, as
had been published by Peligot in 1842. Hence, this was the first process introduced
at Mallinkrodt, near St. Louis, Missouri and at Springfields in the UK, operated by
ICI (later British Nuclear Fuels Limited (BNFL) and then Springfields Fuels Limited
(SFL)). The process required nitric acid dissolution, ether extraction, washing the
ether extraction to remove impurities, reextracting with larger volumes of water, and
conversion of the purified uranyl nitrate to uranium trioxide. With wartime urgency,
production of purified uranium at more than 1 t/day was achieved within 2 months
of starting the project (Mallinkrodt, 1962).
With further research, it was determined that tributyl phosphate (TBP) in kero-
sene was a superior extractant, principally for safety reasons. This TBP system was
established in a large production facility, called the Feed Materials Production
Centre operated by National Lead at Fernald, just outside Cincinnati, Ohio. In
1956, essentially a duplicate of the Fernald solvent extraction plant was built by
Eldorado in Port Hope, Ontario. Likewise, similar commercial TBP solvent extrac-
tion processes were established in the UK and France.
The US Atomic Energy Commission (AEC) had installed UF6 conversion capa-
bility at its US enrichment sites, but later decided this activity could be handled by
the private sector. So, in late 1955, the AEC solicited proposals for the production
of refined uranium products by privately owned and operated facilities. This led to
support of a proposal from Allied Chemical Corporation as this was expected
to provide the lowest overall cost. Hence, a plant was constructed in Metropolis,
Illinois for production of refined UF6, through conversion of concentrates and puri-
fication by distillation (Ruch et al., 1960).
With increasing demands at that time, the Kerr-McGee Corporation (later called
Sequoyah Fuels) completed construction of a “wet process” conversion facility in
Gore, Oklahoma, with capacity of 5000 tU/a in 1970, and later increased that capac-
ity, but this facility is no longer operational.
The wet process is now exclusively used on a large scale in Canada, France, Russia,
and China, and in smaller plants in various other countries, such as Brazil, Pakistan,
and Iran. In Russia, the government enrichment and conversion company TVEL, a
subsidiary of Atomenergoprom, is consolidating its UF6 production from Angarsk to
Seversk. Russian UF6 production has historically received UF4 feedstock from the JSC
Chepetsky Mechanical Plant and concentrates from domestic and foreign mines.
The second commercial process (dry process) for nuclear-grade UF6 is achieved
by production of impure UF6 from concentrates, again using the steps shown by the
three preceding chemical equations, but then achieving the purification by distilla-
tion of the impure UF6. This process is currently only used for natural uranium by
Honeywell in Metropolis, Illinois, and the product is marketed through Converdyn,
a 50:50 partnership of Honeywell and General Atomics. This plant was previously
owned by Allied Chemical and then Allied Signal, prior to Honeywell.
301Conversion of natural uranium
11.3.3 Uranium oxide concentrate refining/conversionto UF6 by wet process
The essential steps in refining by solvent extraction (wet process) are dissolution in
nitric acid, solvent extraction, concentration and thermal decomposition to nuclear-
pure UO3 powder. This process is well documented (Harrrington and Ruele, 1959;
Ashbrook and Smart, 1980; Ashbrook, 1982; Page, 1986).
The wet process is used in the current operations at Cameco Corp. in Canada,
(formerly Eldorado Nuclear) in Blind River (Schisler et al., 1986; Astles and
Green, 1998) and Port Hope, Canada, and by Comurhex/AREVA in France
Comurhex/AREVA in Malvesi and Pierrelatte, France (Delannoy and Faron, 1982a,b).
These are discussed in further detail next. Both operations receive concentrates
from around the world. The concentrates are expected to meet the specifications
that are based on the generic ASTM specification for such concentrates. A sliding
scale of penalties may be applied in treatment pricing for deviations from the
specifications.
11.3.3.1 Cameco uranium oxide concentrate refining to UO3
The flowsheet outlining the Cameco refining process is shown in Fig. 11.1, and in
its current form, was described by Cameco in 2011 in regulatory documentation
(Cameco, 2012).
Blind river refining process
Mine ore
Mill
Uranium ore
concentrate
(U3O8) to
refinery
Nitric acid
3-Stage digestion
Water
3-Stage
evaporation
Aqueous
uranium
extract
Concentrated
uranyl nitrate
hexahydrate
solution
Uranyl nitrate solution
Uranyl nitrate solution
Steam
Vapor
recovery
Vapor
recovery
Vapor
recovery
Solvent extraction
Recycle to digestion
TBP solvent
Raffinate from extraction
Condenser water vapor
Raffinate evaporators
Acid
concentrator
Nitric acid recycle
to digestion
Acid
absorber
Nitrogen
oxides to
acid
recovery
Denitration
reactor
UO3 powder
shipped to Port Hope
Conversion Plant
Calciner Drum dryer
Calc
ined b
ypro
duct re
cycle
to m
ill
Figure 11.1 Outline of Cameco’s uranium oxide concentrate refining process.
302 Uranium for Nuclear Power
UOCs are sampled and split to provide required samples for the producer, the
original mill, and for possible umpire analyses should any dispute arise. The ore
concentrates are dissolved in nitric acid to produce a uranyl nitrate solution. Small
amounts of scrap uranium-bearing materials are recycled to this digestion step.
Impurities are removed from this solution using solvent extraction by TBP in a
kerosene-type diluent. This has three principal steps: extraction, scrubbing (impurity
washing), and reextraction (also called stripping). The extraction chemistry is
described by the equation:
UO212 ðaqÞ1 2NO2
3 ðaqÞ1 2TBP ðorgÞ ! UO2ðNO3Þ2:2TBP ðorgÞ
From the equation, it can be easily understood that higher nitrate concentrations
push the equilibrium to the right. Hence, this encourages the uranium extraction in
the first stage. Reextraction (also known as stripping) is achieved by promoting the
reverse reaction by using little or no nitrate in the reextraction feed water.
Intermediate scrubbing uses a relatively small amount of feed water to remove less
well-extracted impurities, with minimum removal of uranium. The resulting scrub
solution is recycled to extraction for recovery of the uranium content. The stream
of impurities in aqueous solution from the extraction step is known as raffinate.
This is evaporated to dryness and because the solids still contain uranium values,
they are recycled as a feed to uranium mills.
The purified uranyl nitrate solution from reextraction is often given the historical
name of “OK liquor.” This solution is heated and concentrated, producing a nuclear-
grade uranyl nitrate hexahydrate (UNH). The UNH is a yellow crystalline solid but,
on warming, readily dissolves in its own water of hydration, and hence moves as a
liquid between unit operations. This UNH is thermally decomposed in heated and
stirred pots to form UO3 powder. The UO3 is stored in specially designed bulk con-
tainers and these are shipped to Cameco’s conversion facility in Port Hope, but there
is also ability to fill drums for customers who may require nuclear-grade UO3.
Water is condensed from the UNH and raffinate evaporation steps, and this can
be recycled to provide water for the solvent extraction steps. Nitrogen oxide fumes
are generated from various steps in the process, and these are absorbed and concen-
trated to provide nitric acid that is recycled to digestion. (Ozberk, 2011).
11.3.3.2 Comurhex refining and conversion to UF4
The flowsheet published for the Comurhex refining process is shown in Fig. 11.2.
The process involves TBP solvent extraction with column contactors similar to
the Cameco operation.
For Comurhex, a more reactive UO3 is required for the downstream gas/solid
reactions, so their published process involves evaporation and then precipitation
with ammonia to make an ammonium diuranate (ADU), which is dried and calcined
to UO3. A byproduct ammonium nitrate solution is also generated.
ðNH4Þ2U2O7 ! 2UO3 1 2NH3 1H2O
303Conversion of natural uranium
Smaller amounts of nitrogen oxide fumes from this process are collected and
neutralized for disposal.
The direct route has the benefit of avoiding a byproduct ammonium nitrate
stream requiring disposition.
11.3.3.3 Process byproducts and environmental aspects
Generally byproducts of processing result in some materials with low residual ura-
nium concentrations, but still of some value. Typically, the solids are dissolved or
leached in nitric acid to recover the uranium into solution. For these types of dilute
solutions, the uranium is concentrated by precipitation with ammonia or sodium
hydroxide. The impure concentrate can then be collected and recycled with mill con-
centrates to re-enter the main process stream. An alternative strategy is to use a weak
liquor solvent extraction column. The uranium-rich organic stream can be fed into
the main extraction column at a point equivalent to the same uranium concentration.
In the solvent extraction process, the organics are slowly oxidized and the result-
ing generally acidic degradation products are removed in a solvent treatment step.
This process is generally known as solvent regeneration. This process results in a
regeneration product high in the oxidation products as well as some residual ura-
nium and this can be recycled as a mill feed for uranium recovery.
11.3.4 Uranium oxide conversion to UF6 after refining
11.3.4.1 Reduction to UO2
Reduction with hydrogen at high temperature is carried out in fluid beds at
Cameco’s Port Hope plant. Some increase in surface area is generally preferred
Uranium
concentrates
Sampling
Precipitation
Filtration
DryingCalcination Compaction
UO3 Storage
Solvent centrifugation
Demineralized
Water
HNO3
NH3
UO3
Drums emptying
Dissolution
Extraction Reextraction
Concentration
Reduction
NH3
UF4
Storage
UF4
Shipping
UF4
HF
FluoridationFertilizerWashing
Impure
uranyl nitrate
Decantationconcentration
vapor
Mothers watersammonium nitrate
Impurities
Solvent
regeneration Storageuranylnitrate
Calcination
Impurities
Na2CO
3
Figure 11.2 Outline of Comurhex refining process.
304 Uranium for Nuclear Power
prior to reduction. At Cameco, this is achieved by pulverizing. In the reduction, suf-
ficient fluidizing gas is needed which may be achieved through recycling of hydro-
gen or use of inert gas with the hydrogen.
Gas/solid contact in a moving bed was used at National Lead and in the early
days of Port Hope plant. This method remains in use with Comurhex at Malvesi.
UO3 is compacted and broken into suitable size pieces for entry into the top of the
moving bed or lit courant (LC) reactor. The UO3 moves by gravity into the high-
temperature reduction zone of the reactor. Reduction is achieved by introduction of
ammonia at the base of the reactor, where it dissociates into hydrogen and nitrogen,
catalyzed by uranium oxide produced in the reduction. The solids then pass through
a cooling zone and on to hydrofluorination.
11.3.4.2 Hydrofluorination to UF4
At Comurhex, conversion to UF4 is in the same type of LC reactors, where liquid
anhydrous HF is fed through preheaters for vaporization, and into the UF4 dis-
charge end of the reactor. The gas moves counter-current to the uranium, so that
most reaction to UF4 takes place in the vertical section of the reactor where the
temperature is relatively high. The upper portion, which mostly contains UO2,
acts as a clean-up to maximize use of HF (Delannoy and Faron, 1982a,b). The
UF4 product is transported from the Malvesi site to the Pierrelatte site for subse-
quent conversion to UF6.
Both Comurhex and SFL in the past have historically diverted some best qual-
ity UF4 to make metal, but both sites have ceased making metal. However, for
Comurhex, it was the initial focus on UF4 for metal that led to geographical sepa-
ration of the UF4 production at Malvesi and UF6 production at Pierrelatte
(Fig. 11.3).
Cameco chose to use a wet hydrofluorination process, which they call “wet
way” (Lenahan, 1982). In this process, pulverized UO2 powder is fed into stirred
slurry reactors with water and hydrofluoric acid. On drying the UF4, most of the
vaporized hydrofluoric acid is recycled from scrubbers and this results in very
effective use of the HF. The downside of this wet process is that subsequent drying
inevitably results in some reversion to oxide and purity that would be unsuitable for
metal production. However, this is no longer a significant market consideration;
rather the focus is on processes with least environmental impact.
11.3.4.3 Fluorination to UF6
Commercial conversion of UF4 to UF6 is carried out by reaction with fluorine gas.
Most commonly, this is achieved in vertical shaft reactors where the fluorine and
UF4 powders are mixed together at the top of the shaft and the reaction is so exo-
thermic and fast that the reaction is largely complete during the time of fall of the
powder down the shaft. Due to the highly exothermic nature of the reaction, the
reactor temperature is naturally very high. This type of reactor is used at Angarsk,
Port Hope, and Pierrelatte, but with some differences in dimensions and number of
reactors.
305Conversion of natural uranium
Both Cameco and Comurhex use two reactors in series; in the first reactor,
there is excess fluorine encouraging complete conversion of all UF4 and in the
second reactor there is an excess of UF4 thereby ensuring complete use of the
expensive fluorine gas (“clean-up” reactors). Several reactors of each type operate
in parallel. The remaining UF4 from the clean-up reactors is recycled as part of
the feed to the primary reactors. Solids collected at the bottom of the primary
reactors, often called primary ash, are recycled to the primary reactor feed, while
solids subsequently collected from the UF6 gas stream are a secondary ash and
treated by appropriate uranium recovery methods, possibly as mill or refinery
feeds, but only after ageing for some months to reduce radiation fields associated
with uranium daughters that are no longer shielded by uranium. Treatment to
avoid HF fumes, possibly using basic compounds additions, is also necessary for
safe shipping of this ash.
Gaseous UF6 is condensed in cold traps with internal heat exchange pipes that
contain a refrigerant. Typically when the first condenser is full, the second con-
denser is moved to the primary position and an empty condenser added in the sec-
ondary position. After a trap is filled in the primary position, it is heated to about
95�C and under the small overpressure, the UF6 liquefies and can be drained off
into storage/transport cylinders via remotely operated filling systems.
PROCESS OVERVIEW
UO3 powder
UO3
powder
UF6
liquid
UF6 cylinders for export
(UF6 solid)
UF6
gas
UO2
powder
Anhydrous
hydrofluoric acid (HF)
Off gas
Off gas
Hydrogen gas (H2)
Fluorine gas (F2)
Ash
ACID RECOVERY
SCRUBBERS
ACID RECOVERY
SCRUBBERS
PULVERIZER
FLUID BED REACTORS
STIRRED REACTORS
COLD TRAPS FLAME REACTORS
DRUM DRYERS CALCINERS
Aqueous
hydrofluoric acid
UF4
slurry
UF4
powder
UF4
powder
Figure 11.3 Outline of Cameco’s UF6 conversion process.
306 Uranium for Nuclear Power
11.3.5 Uranium oxide concentrate conversion to UF6prior to refining
The process now uniquely used at Honeywell in Metropolis, Illinois, is the conver-
sion of UOC to impure UF6 and then purification by distillation (Lawroski et al.,
1958; Ruch et al., 1960; Bishop and Hanson, 1982). Ore concentrates are sampled
through a falling steam system, where a 1% sample is further split to give about
1500 g/batch. This is typically divided into six 150-g samples, two for Honeywell’s
analysis, two for the originating mill or designee and two are retained for umpire
use as required.
In the Honeywell process, concentrates can be passed through a calciner to
remove carbonates, water, organics, and other volatiles. The calcined material
is blended and agglomerated to optimize particle size for fluid-bed operation.
Feeds high in sodium are dissolved in sulfuric acid, ammonia added, the slurry
settled, filtered, resuspended in water, again settled, and filtered. The resulting
solids are returned to the raw concentrates calciner. The operation of the
calciner is considered critical to good operation of the fluid-bed reactors
downstream.
The chemistry for UF6 production is essentially the same as described previ-
ously, and this is followed by distillation for purification. The gas/solid reactors
used in each stage are fluidized bed reactors. For the conversion of UF4 to UF6,an inert bed of CaF2 is used in the fluid bed because a bed of UF4 can
easily lead to fusion of bed material due to the high-energy release by this
fluorination reaction and relatively low melting temperatures of UF4 and interme-
diate fluorides.
The process flowsheet for the Honeywell plant, Metropolis is presented in
Fig. 11.4 Uranium 2000 short course notes, Chapter 10, presented by Tom Rice,
METSOC/CIM.
After the UF6 gas leaves the fluorination step, it is filtered and then condensed
in specially designed heat exchangers. When full, the collection units go through a
cycle of heating and melting the UF6, which is sent as liquid under a small over-
pressure to distillation. There are two distillation units in series, first the low boiling
impurities are removed from the top of the first distillation column, and then high
boiling impurities are removed at the bottom of the second column. Heels from
UF6 process have high radiation fields and are stored temporarily in lead shielded
drums. Liquid UF6 from the distillation product condenser is drained into transport
cylinders. The UF6 is sampled during filling and the cylinders weighed and
removed to the storage area where they are allowed to sit for 5 days to solidify prior
to further handling.
Byproduct high uranium materials are sent to a recovery unit and returned to the
UOC calciner. Potassium hydroxide (KOH) scrub liquors are regenerated with lime.
Weak HF solutions are also neutralized with lime. Other solutions mostly water
potentially contaminated with insoluble uranium are passed through a series of set-
tling ponds prior to discharge.
307Conversion of natural uranium
FROM
DRUM DUMPING
&
SODIUM REMOVAL
SYSTEM
AGGLOMERATION
SIZING
TO
REDUCTOR
FILTER
REDUCTOR
FROM
ORE
PREP
FEED
HOPPER
PREHEATER
DISSOCIATOR
INCINERATOR
NH3
UO2
TO
GREEN
SALT
CALCINATION
PREPARED FEED
REDUCTION
NH3
VAPORIZERS
Figure 11.4 Outline of Honeywell conversion and refining process.
308 Uranium for Nuclear Power
SODIUM REMOVAL
H2SO4
HIGH SODIUM
ORE CONCENTRATES
AMMONIA
SETTLER
SETTLER
NEUTRALIZER
DIGESTER
FILTER
FILTER
TO CALCINER
WATER
RESLURRY
GREEN SALTHF
SCRUBBERS
TO
EPF
GREEN
SALT
HOPPER
TO
FLUORINATION
ELEVATOR
PREHEATER
HF
VAPORIZERS
HF
FILTERS
FROM
REDUCTION
PRIMARY
HYDRO
FLUORINATOR
FILTERS
SECONDARY
HYDRO
FLUORINATOR
Figure 11.4 (Continued)
309Conversion of natural uranium
DISTILLATIONFROM FLUORINATION
TO COLD
TRAPS
LOW
BOILER
CONDENSERS
PRODUCT
CONDENSER
HIGH
BOILING
SYSTEM
UF6 PRODUCT
FILLING AND
SHIPPINGHIGH
BOILER
POT
LOW
BOILING
SYSTEM
STILL
FEED
TANKS
LOW
BOILER
POT
VAPORIZER
FLUORINATION
FILTERS
FLUORINE MUDS TO
PROCESSING
BLOWER
TO
ATMOSPHERE
TO
DISTILLATION
COLD
TRAPS
FROM
GREEN
SALT
UF4
MILL
UF4
BLENDER
EXIT
SCRUBBERS
FLUORINATOR
Figure 11.4 (Continued)
310 Uranium for Nuclear Power
11.3.6 UF6 product
Verifying of the quality of the UF6 product is an important step. For Cameco, this
is achieved by remelting of UF6 in representative cylinders, and withdrawing sam-
ples for analysis on specification elements. Comurhex prefers to have an interme-
diate liquid UF6 hold tank that is large enough to fill one cylinder completely and
allow for homogeneity and settling of fine solids. The liquid is then decanted to
the transport cylinders. Product quality assurance procedures must be rigorously
maintained, although this does not mean every cylinder must be sampled, only
that the sampling procedures are such that customers are assured of product
quality.
The containment of UF6 requires particular care, primarily due to three factors:
1. UF6 expands by about 36% when solid is melted to liquid so it is essential that cylinders
not be overfilled. If a mistake is made, it can result in bursting of a transportation
cylinder.
2. UF6 reacts immediately with any moisture to yield HF and a very visible cloud of
UO2F2. This acts as a warning of leaking and it is now common that cylinder filling
areas have the ability to freeze out any possible liquid UF6 with solid carbon dioxide
sprays.
3. UF6 is a very strong oxidizer and should any hydrocarbon oils or greases mix with the
UF6, there is an explosion risk.
All producers take great care in the handling of UF6, particularly as liquid; in
some cases, UF6 is only moved as a gas or as a solid, including collection of the
solid in tote bins for transfer to an enrichment plant. The valving of containers with
larger openings for solid and gas transfer adds some risk as well. All fluoride-
containing gas streams are typically scrubbed with aqueous KOH.
11.3.6.1 Fluorine production
Fluorine is generated in greater amounts for UF6 production than any other product,
and is a very expensive and hazardous part of this production. Hence, fluorine pro-
duction technology is very important to economic UF6 production. Fluorine is
produced by electrolysis of the fused salt, KF.2HF, at 85�90�C in mild steel cells,
using amorphous carbon anodes and mild steel cathodes. The cells operate continu-
ously with HF added appropriately by monitoring electrolytes levels. Water cooling
is also needed. More in-depth information of this subject is available in references
(Ellis and May, 1986; Shia, 1995; Slesser and Schram, 1951).
Fluorine leaving the cells contains about 7% HF, which may be used as is
(Cameco) or the HF removed by reaction with sodium fluoride. With no HF
removal, the HF passes through UF6 collection and is usually trapped in very cold
secondary traps, allowing it to recycle into the hydrofluorination process.
The byproduct hydrogen stream also contains about 7% HF. Typically the HF
from the hydrogen stream is removed by cold traps and scrubbers before the hydro-
gen gas is vented to atmosphere.
311Conversion of natural uranium
11.3.6.2 Process comparisons
AEC chose to promote the distillation process based on cost, however the rest of
converters have chosen the solvent extraction process. One published report (Fryer,
1968) suggests the economics are not much different between the two processes
and other operating parameters drive the choice between the two options. These
factors include:
1. The solvent extraction process is believed by most to be more tolerant of higher levels of
impurities in the feed concentrate.
2. Sodium and potassium present in the feed concentrate tend to make nonvolatile salts such
as Na2UF8, reducing uranium recovery and further lowering the sintering temperature of
the bed in the fluid bed, causing lumping. These alkalis metals can be pretreated for
removal but the solvent extraction process has the advantage of rejecting these elements
without an additional step and increased wastes, if uranium oxide concentrates are
expected to be routinely high in concentrations of the alkali metals.
3. Some organics and other impurities can promote emulsions in solvent extraction, whereas
these impurities would not be as problematic in the fluorination/distillation process.
4. The presence of impurities during fluorination increases consumption of expensive HF
and F2 in the distillation process, so higher levels of most impurities support use of the
solvent extraction process.
5. In the distillation process, the presence of impurities, such as S and As, in the initial
reduction, produce higher concentrations of noxious gases (H2S and AsH3), which must
be scrubbed at extra cost. The distillation process also produces small amounts of impu-
rity streams rich in MoF6 and VOF3, which are difficult to dispose.
6. Distillation does have the advantage that if corrosion products from process equipment
add to product impurity levels, then impurity removal near the end of processing has
advantage.
7. The nuclear-pure UO3 from the solvent extraction process may have other commercial
value. For example, Cameco takes some of this stream to make sinterable natural UO2
powder for fuel manufacturing mainly for CANDU reactors and making blanket fuels for
light water reactors.
8. The nuclear-pure UF4 from the solvent extraction process may have other commercial value.
For example, SFL took some of this stream for uranium metal production for Magnox reac-
tor fuel, and Comurhex also took advantage of this feature in the distant past.
9. The distillation process may be preferable for enriched uranium as the absence of water
reduces the risk of criticality.
11.3.7 Safety
A major safety concern for the production of UF6 is the safe handling of anhydrous
HF. Different approaches can be taken to achieve acceptable levels of safety. At
Malvesi, HF storage and handling of anhydrous hydrogen fluoride (AHF) is entirely
outside. At Honeywell, the HF is unloaded from outside transport containers and
significant HF releases addressed by a curtain of water cannon, as has become stan-
dard practice at HF production facilities in North America. At Cameco, the
approach is to ensure all unloading is within the production facility, where
312 Uranium for Nuclear Power
collection of any fumes is ensured and discharge only after passage through the
plant scrubbers. The system is designed so that even if there were significant leak-
age the HF would be contained within the building. (Clark and Kennedy, 2010).
Within plants, both HF and UF6 present significant gaseous hazards. Where pos-
sible, systems operate at below atmospheric pressure so that any leak would be
inward, not outward. However, this is not always possible particularly where liquid
UF6 is present because this requires a slight overpressure. Also, the more progres-
sive plant designs have ability to segregate operating areas so fume collection capa-
bilities can be focused in specific areas when considered necessary. After multiple
dust collection and scrubbing systems, ideally there is one final discharge point at
high level where there is online monitoring of HF and U emissions.
There is typically wide use of sensors for uranium dust, fluorine, hydrogen fluo-
ride and closed circuit TV, with outputs to both a central control room and localized
control rooms where necessary, for example, HF unloading, cylinder filling.
Fluorine detection is usually through UV analyzers and stack fluoride emissions by
absorption and ion-selective electrodes. Fine uranium dust levels are usually mea-
sured by filters that are taken off line as frequently as appropriate and are therefore
not truly continuous for control room information. Even very small leaks of UF6immediately form very visible clouds of UO2F2, which may be detected by smoke
detectors or by visual appearance on camera. Thus, although unlikely hazardous
releases are quickly recognized and addressed with minimum exposures to operat-
ing staff.
Large parts of the systems are automated and use is made of remote control sys-
tems and closed control rooms with closed circuit television systems to minimize
any chance of gas exposure to operators.
11.3.8 Environmental
Process byproducts, even with low uranium contamination are increasingly difficult
to dispose of, mainly due to negative public perceptions. One would think, because
very low uranium levels are already widely present in the environment it would not
be a big issue. However, this is a reality of the industry and in recent years has
probably been the greatest driver for process changes to ensure minimizing uranium
emissions.
In general, sites scrub final off-gases from hydrofluorination and fluorination
with KOH solution. This is preferred over NaOH solution due to the relatively poor
solubility of NaF, over more soluble KF, produced from the scrubbing of HF.
Potassium diuranate may be separated off and recycled to refining or milling
operation. The exhausted scrub solution may be evaporated to dryness to yield a
uranium contaminated KF/K2CO3 byproduct, or treated with lime to regenerate the
KOH and produce uranium containing CaF2/CaCO3/Ca(OH)2 byproduct. Either
product may be recycled to a mill for uranium recovery, although some have cho-
sen to simply send this for disposal.
313Conversion of natural uranium
11.4 Current status of the conversion industry
As noted previously, the commercial UF6 conversion plants, with best capacity esti-
mates from published data, are as follows (WNA, 2014; IAEA, 2012, 2009):
Operator Country Location Nominal capacity
(tU/a)
Cameco Canada Port Hope, ON 12,500
Comurhex (AREVA) France Malvesi and Pierrelatte 15,000
Rosatom Russia Seversk 12,500
CNNC China Lanzhou 5,000
Honeywell (Converdyn) United States Metropolis, IL 15,000
Although nominal capacities for UF6 production are given, the current low
demands for conversion means that most plants are operating well below, and some
near 50% of these nominal capacities. Also, nominal capacities for refined UO3
may be substantially higher as some of this intermediate product may be diverted
for production of other products such as uranium metal and ceramic-grade UO2.
Most significantly for Cameco, Blind River refinery has almost twice the capacity
(24,000 tU as UO3) relative to UF6 production capacity due to requirements of pro-
ducing other products during its history.
Major changes are expected in France with a new plant to be built in Pierrelatte
to accommodate all refining and conversion operations with some eventual increase
in overall capacity.
It is of note that Russia has historically produced UF6 at two sites but has enrich-
ment plants at four sites. Currently, UF6 is produced at one site at Seversk and the
Angarsk plant was shut down in 2014. It is believed partially enriched UF6 is
shipped as feed to the sites without UF6 production capacity.
The China National Nuclear Corporation (CNNC) production is estimated to be
B5000 tU/a. It is expected to continue increasing to meet domestic enrichment
requirements.
There was commercial capacity of about 6000 tU/a at Springfields in the UK,
(Ozberk, 2011) but this was closed in 2014. Construction of a future commercial
capacity of about 6000 tU/a was publicized by Kazakhstan from a joint venture of
Kazatomprom and Cameco, but timing remains uncertain. Kazakhstan appears com-
mitted to become an UF6 producer adding value to its large uranium production.
There are also a number of small-scale or “pilot” plants for internal national sup-
plies of UF6, sometimes associated with nuclear weapons programs and therefore
known with less certainty, notably:
Operator Country Location Capacity (tU/a)
AEOI Iran Isfahan B200
BARC India Trombay Unknown
CNEA Argentina Rio Negro Unknown
IPEN Brazil Sao Paulo B100
North Korea Unknown
CPC Pakistan Dera Ghazi Khan Unknown
314 Uranium for Nuclear Power
These operations are all believed to use the wet process refining; certainly, the
deliveries of TBP have been used as indicators of probable uranium refining activ-
ity. However, there are few published details on these operations.
AEOI of Iran started the plant in Isfahan in 2004 and this plant is under IAEA
safeguards. Production is tracked by various organizations, such as IAEA.
India has an unsafeguarded plant called the Uranium Metal Plant at the Bhabha
Atomic Research Center (BARC), which converts yellowcake to UF4 and then
either to metal or UF6. There is also a larger refining capacity as some refined prod-
uct is diverted to make natural UO2 for their heavy water moderated reactors.
The National Atomic Energy Commission of Argentina (CNEA) has facilities at
the Complejo Tecnologico Pilcaniyeu site, in the Rio Negro district, which includes
a UF6 conversion plant.
The Brazilian pilot-scale process is largely based on French technology and uses
the ADU route to UO3.
The Pakistan Chemical Plants Complex (CPC) converts yellowcake to UF6, and also
enriched UF6 to metal and depleted UF6 to metal. This undeclared and unsafeguarded
nuclear site was built in the 1970s and early 1980s, and continues to be upgraded.
11.5 Factors that impact converters
A very significant issue is that the two North American converters are economically
separate from enrichment, unlike all other such operations around the world. Thus,
they work to a different business model, to justify their activity independent of
enrichment revenue in this step of the fuel cycle. This is a challenge as historically
refining/conversion to UF6 has generated relatively little revenue and, on average,
little profit. The cost structure is improved somewhat by increased capacity, as long
as that capacity is well used, that is over 80%. Of course, in recent years this has
not been achieved and is a significant part of the reasons why the smaller converter
in the UK, Springfields Fuels Ltd ceased production recently.
Government operations tend to act to ensure conversion is available for their
national enrichment programs, independent of the economics of this separate step.
Probably in recent years, this type of operation is most evident in China, for as the
national enrichment capacity has increased, so has the national capacity for conver-
sion to UF6.
Other risks must be considered along with the politics and economics of the
business. Of note are:
Chemical contact risk: Several hazardous chemicals are essential parts of refin-
ing conversion. These include hydrogen fluoride, fluorine, UF6, nitric acid, sol-
vents, and TBP. Unintended contacts with employees or releases outside plant
boundaries can have very negative impacts. Equipment and operating conditions
are designed to minimize risks but some small risk remains. When upsets occur,
there are many barriers to maintain containment, but some unplanned releases to
the environment still occur, although are becoming less through proactive more
stringent safety programs.
315Conversion of natural uranium
Chemical supply risk: Hydrogen fluoride is a strategically important chemical its
major use is as refrigerants, and as world affluence grows so does the demand for
air conditioning, and hence HF. World supplies are dominated by China, and China
does not hesitate to use supply dominance to protect home industries. Hence,
demand/supply risks for HF can be expected to increase. The extremely toxic nature
of HF has also caused some jurisdictions to consider banning the chemical, and
acceptance of its transport through communities can be expected to become more
challenging. Organic phosphates, such as TBP, are considered a cancer risk,
although there is no evidence of such a problem for TBP in uranium refining. Still
supply disruptions must be considered a risk.
Radiological risk: Radiation doses are generally well below regulatory standards
but still some risk is associated with the uranium daughters when procedures,
shielding, and containment are not properly maintained. It is a feature of the indus-
try that when releases might have impact outside a facility, the public express pri-
mary concern over radiation, although chemical impacts are likely to be greater. In
turn, regulators usually target the focus of public concern.
Recycle risk: Refining and conversion result in byproducts, containing impuri-
ties but usually some residual levels of uranium that still have value. Recycle to
mix such material with uranium mill feeds is usually possible, but the transport of
such materials, which have at times wrongly been labelled as waste, usually
attracts public and regulatory interest, resulting is some uncertainties in availabil-
ity of these routes.
Waste risk: Contaminated equipment that is beyond useful lifetimes needs to be
cleaned and residues disposed. Some other materials such as insulation, plastics,
wood, and paper can be contaminated and not easily cleaned. Incineration may be
useful in allowing some uranium recovery but such incinerators are no longer easily
licensed. Although very small there is inevitably some waste, and disposal can be
difficult. Some historic practices of simply storing are now generally a regulatory
issue, demanding action.
Inventory risk: Refining/conversion sites have historically held the uranium
inventories for customers and these have been as large as the feed/production for
1 year. The value of such an inventory can approach 1 billion dollars. Fortunately
the weight is very substantial so theft is not likely, concerns are greater that in
some way the inventory could be stranded by strikes, plant upset, transport or reg-
ulatory issues. For these reasons, there is a tendency in recent years for lower
inventories to be held at refining/conversion sites and more to be held at enrich-
ment sites.
11.6 Conversion to UO2
Natural uranium is also converted to ceramic-grade UO2 for production of reactor
fuels. The demand is mainly for CANDU reactors, but some use also exists for
pressurized water reactor (PWR) and boiling water reactor (BWR) blanket fuels.
316 Uranium for Nuclear Power
Production may be through natural UF6 just as enriched UO2 is produced
through enriched UF6, as discussed further in Chapter 13 “Nuclear fuel fabrica-
tion.” A more cost-effective process is through conversion of natural nuclear-
grade UO3. This process involves precipitation of ADU from ammonium nitrate
solution using ammonia, separation and drying of the ADU and then thermal
reduction with hydrogen at high temperature to yield the UO2, as it is illustrated
in the flowsheet published by Cameco Corp., shown in Figure 11.5.
In Cameco’s current operation (Itzkovitch and Zawidski, 1985; Kwong and
Kuchurean, 1997; Ozberk, 2007), UO3 from the refining at Blind River as
described previously, is dissolved by nitric acid to give the feed uranyl nitrate
solution. Fume collection is provided at the tank to prevent nitric acid vapors
from escaping into the plant. In earlier days of Cameco operations, when refining
and conversion were on the same site, the uranyl nitrate solution could be taken
after solvent extraction/partial evaporation and prior to denitration, but now
because UO3 is the form used for uranium transport between distant sites the dis-
solution step is needed.
Prior to precipitation, the concentrated uranyl nitrate is diluted and then ammo-
nia added to yield an ADU precipitate. These solids are collected, and reduced to
UO2 powder in externally heated reactors with counter-current flow of hydrogen.
Conditions are set to produce a powder with an O:U atomic ratio of 2.05�2.18,
Nitric acid
THE PROCESS
DISSOLUTION TANK
PRECIPITATION TANKS CENTRIFUGES
DRYER
SCRUBBER FILTER
ROTARY KILNS
BLENDER
Uranyl
nitrate
Aqueous ammonia
Ammonium
diuranate
Off-gas Off-gas
UO2
UO2
H2
Ammonium
diuranate slurryAmmonium nitrate
for fertilizer
Natural UO2 for
light water reactorsNatural UO2 in drums
for candu fuel reactors
UO3 powder
Figure 11.5 Outline of Cameco’s UO2 conversion process.
317Conversion of natural uranium
which can be handled without difficulty in air. The powder is then blended in
batches, packaged and shipped to fuel manufacturers.
While Canada is the primary commercial producer of such natural UO2, there
are other small plants in Argentina, China, and Romania where national self-
sufficiency is a priority.
11.7 Potential future developments
Technology, processing, and conversion are not likely to change significantly in the
near future. No product other than UF6 is likely to be in demand for enrichment
and the overall demand will only increase slowly. Indeed as enrichers increase
enriching efficiency and more high-235U tails are used as feed, then there is reduced
demand for conversion.
Potential increasing demand for conversion, resulting from the expansion of
nuclear programs in Asia, particularly China and India, will be addressed by more
rapid expansion of capacities as national control takes priority over economic con-
siderations. This will result in little chance of increasing demand for conversion in
the Western world and no substantial increases in price due to the competition
inevitable from demands that do not cause plants to operate near capacity. The
small profit margin results in limited research on process improvements, and that
which does happen will address the emission and waste issues of public concern, as
required to ensure continued public and regulatory acceptance of these operations,
and security of supply for customers.
In short, limited changes can be expected in processing technologies, except for
some production growth in Asia in the near future.
References
Ashbrook, A.W., 1982. Uranium refining and conversion practice in the western world: an
overview. In: Proceedings of Uranium ’82, Toronto, Canada, CIM, 1982.
Ashbrook, A.W., Smart, B.C., 1980. A review and update of refining practice in Canada. In:
Production of Yellow Cake and Uranium Fluorides, 5�9 June 1979, Paris, Published
IAEA proceedings, Vienna, 1980, pp. 261�287.
Astles, C.R., Green, C.A., 1998. New process for natural uranium by-product preparation in
uranium refining operations. In: Proceedings of Pacific Nuclear Conference, Toronto,
Naturally occurring uranium comprises about 99.29% of the isotope 238U (146 neu-
trons, 92 protons) and 0.71% of the isotope 235U (143 neutrons, 92 protons).238U is mildly radioactive, undergoing alpha decay to 234Th, but it is not fissile,
that is, it cannot be used on its own to sustain a nuclear chain reaction.235U, on the other hand, is a less stable isotope that undergoes fission when a
thermal neutron collides with the nucleus. More neutrons are emitted and, in the
right quantity and concentration, 235U is capable of sustaining a nuclear chain reac-
tion; for example, as is required for a power generating nuclear reactor.
Uranium enrichment is the name given to any isotope separation process applied
to natural uranium resulting in a product in which the concentration of 235U is
increased above the natural concentration of 0.71%.
The capacity of any equipment or installation to enrich uranium in measured in
separative work units (SWUs).
12.1.2 What is a SWU?
Separative work performed to enrich uranium is defined by a mathematical formula
(see next), but in essence it is a measure of work performed by a process to take a
quantity of feed material at a certain 235U concentration and convert it into a quan-
tity of enriched product with a higher 235U concentration and a balancing quantity
of depleted “tails” with a lower 235U concentration.
An example is shown in Fig. 12.1 where the feed is 0.71% 235U, enriched product
is 4% 235U, and depleted tails are 0.3% 235U.
Precise calculation of SWUs involves differential equations. In ordinary practice,
however, the following equation can be used:
S5Pð2xp2 1Þlnxp
12 xp1 Tð2xt2 1Þln
xt
12 xt2Fð2xf 2 1Þln
xf
12 xf
where:
S is the effort in SWUs,
P is the mass of the product, the enriched uranium,
Uranium for Nuclear Power. DOI: http://dx.doi.org/10.1016/B978-0-08-100307-7.00012-0
One chemical process has been demonstrated to pilot stage but not used. The
French Chemex process exploited a very slight difference between the 235U and238U isotopes valence change behavior in oxidation/reduction using aqueous (III
valence), and organic (IV valence) phases. See also http://world-nuclear.org/info/
12.4.2.1 Atomic Vapour Laser Isotope Separation (AVLIS or SILVAin France)
The feedstock for the AVLIS process is uranium metal. The metal is heated to form
a vapor. Laser light at a very specific frequency is then directed at the vapor such
that 235U atoms only are ionized by the ejection of an electron. Positively charged235U1 ions are attracted to a negatively charged collector plate. The 238U atoms in
the vapor are not ionized by the impact of the laser light and pass unaffected
through the process (Figs. 12.20 and 12.21).
Like all laser isotope separation processes, AVLIS offers the possibility of
lower-energy inputs, lower capital costs and high separation factors; but after
(–) Charge
collector
Uranium vapor
flow
Tailstream
Laser
238U235Ulonized 235U
Figure 12.20 AVLIS process.
Source: NRC.
Figure 12.21 Picture of AVLIS.
Source: Wikepedia.
341Uranium enrichment
research, not least by URENCO, France, and the United States, these advantages
have not proved commercially realizable to date.
Problems thrown up by research include the complexity and reliability of the
required laser systems and the robustness of collector systems. Furthermore,
because it uses metal feedstock, modification to the global fuel cycle, which is ori-
ented around UF6 as the feedstock for enrichment processes, would be required.
See also http://world-nuclear.org/info/Nuclear-Fuel-Cycle/Conversion-Enrichment-
condense to form dimers, trimers, etc., and tuned lasers to selectively disrupt such
condensation of 235U containing species, thus achieving isotope separation.
For more detail, see Eerkens and Miller (2004).
12.5 Quality control of uranium hexafluoridein enrichment
12.5.1 Feed material
Natural uranium hexafluoride feedstock for enrichment is produced at a number of
conversion facilities around the world. Currently the chief facilities in operation are
Cameco’s Port Hope facility in Canada, Converdyn’s Metropolis facility in Illinois,
United States, AREVA’s Comurhex facilities in Malvesi and Tricastin in France,
and conversion facilities in Russia, being consolidated at Seversk.
All natural uranium hexafluoride for enrichment should conform to the 2011
issued standard:
ASTM C787-11 Standard Specification for Uranium Hexafluoride for Enrichment.
Natural uranium hexafluoride feedstock is largely an undifferentiated commodity
regardless of the plant of origin and enrichment facilities typically accept quality
control documentation, including analysis results, from the supplying converter
without resorting to additional sampling and analysis at the enrichment facility.
12.5.2 Enriched product
Enriched product produced at enrichment facilities, forms the feed material to
nuclear fuel fabrication plants across the world. The quality standard, which must
be met by enriched product, is the 2010 issued standard:
ASTM C996-10 Standard Specification for Uranium Hexafluoride for
Enrichment to less than 5% 235U.
The key measurement for enriched product is of course 235U assay. Depending
on feedstock, it is sometimes important to particularly check the concentrations of
other isotopes such as 232U, 234U, and 236U, to ensure the standard is met. 234U and236U are neutron absorbers that depress the reactivity of nuclear fuel while the 232U
decay chain includes hard gamma emitter 208Tl, which creates additional handling
hazards (see Section 12.7).
A variety of sampling and analytical techniques may be deployed at enrichment
facilities, including online mass spectrometry, product off-take stream sampling,
and analysis, to postproduction homogenization and sampling of full product
cylinders.
Figs. 12.24 and 12.25 show a full product cylinder sampling rig at URENCO.
The full cylinder sampling process is as follows:
The cylinder to be sampled is loaded into an autoclave with four sample bottles
connected. The autoclave is closed and the cylinder is heated to liquefy and
344 Uranium for Nuclear Power
homogenize its contents. One end of the autoclave is raised to allow liquid to flow
out of the cylinder valve into the sample bottles. The cylinder is returned to the hor-
izontal and allowed to cool so that the contents resolidify. The autoclave is then
opened and the sample bottles removed, with one for the enrichment company, one
Figure 12.24 30B cylinder inside sampling rig.
Source: URENCO.
Figure 12.25 Sampling rig in tipped position.
Source: URENCO.
345Uranium enrichment
for the customer, one for the umpire, and one spare. The total sampling cycle time
is about 60 h.
12.5.3 Tails
A key point about quality control of tails is its potential to be re-fed for re-enrichment
some time in the future. In this case, it becomes feed material and needs to conform
to ASTM C787-11, as does natural feedstock.
If tails results from original feed conforming to ASTM C787-11, it is almost cer-
tain that the tails will also meet the specification. Online mass spectrometry and
tails off-take stream sampling and analysis are periodically used to check the qual-
ity of tails.
12.6 Management of tails
The management of tails arising from primary enrichment is an important issue for
enrichment service providers.
Three strategies have been deployed by primary enrichers to manage tails:
1. re-enrichment,
2. deconversion,
3. storage pending re-enrichment, deconversion, or some other use.
12.6.1 Re-enrichment of tails
As discussed in Section 12.2.6.1, 235U assays in tails from primary enrichment are
quite typically in the range 0.2�0.3% and are suitable for refeed to further strip out
useful 235U.
Historically, a significant proportion of enrichment capacity in Russia has been
used for the re-enrichment of tails. Russia has not only re-enriched some of its own
primary tails but it has also re-enriched primary tails from both URENCO and
AREVA enrichment operations. The low 235U assay secondary tails arising from
such re-enrichment operations in Russia have been used to dilute HEU derived
from decommissioning of Russian nuclear weapons as part of the “Megatons to
Megawatts Programme” (see Section 12.2.6.3.1).
Both URENCO and AREVA established conventional enrichment contracts with
Russia in which tails from primary enrichment operations at URENCO and
AREVA plants in Europe were transported to Russia to form feedstock for Russian
plants. In return, URENCO and AREVA received back equivalent natural material
or enriched product. Russia took ownership of the secondary tails.
URENCO’s contract with Russia to re-enrich tails ran from 1996 to 2010. In
total 100,000 tU of URENCO tails from its European plants were processed over
this period.
346 Uranium for Nuclear Power
AREVA’s contract ran from 1999 to 2010 and in total 60,000 tU of AREVA
tails were re-enriched.
Elsewhere, other campaigns of tails re-enrichment have been conducted, the
most recent example being the refeed of approximately 9000 t of “high assay” US
DOE tails through the Paducah gas diffusion plant immediately prior to its closure
in 2013. See also http://www.world-nuclear-news.org/C-Tails_deal_gives_Paducah_
another_year-1605127.html.
12.6.2 Tails deconversion
Another option to manage tails is to de-covert tails uranium hexafluoride back to
uranium dioxide. This releases a valuable resource, hydrofluoric acid for reuse in
chemical processes.
Uranium oxide is the most stable form in which to store tails for the long term
and is also suitable for disposition (eg, back into an original uranium mine), should
that be a future decision.
Deconversion of tails has been carried out in France since 1984. AREVA have
operated four deconversion lines (the “W Plants”), at Tricastin with a capacity of
14,000 tU pa. Over 300,000 tU of tails have been deconverted in AREVA’s facility
since the start of operations. Most of the tails processed to date has been arisings
from AREVA’s now closed George Besse I gas diffusion plant. AREVA has also
deconverted some URENCO tails. In contracts over the period 2003�14, about
46,000 tU of URENCO tails have been treated.
Deconverted tails in the form of U3O8 is stored in 10 t capacity cubic steel con-
tainers known as “DV70s.” Deconverted French tails are stores at Bessines
(Fig. 12.26) and Tricastin. Deconverted URENCO tails are stored by the Dutch
national radioactive materials storage organization, COVRA, at Vlissingen in the
Figure 12.26 Picture of DV70 tails U3O8 containers in store at Bessines.
Reprocessed uranium has been fed to gas diffusion plants in the past. Due to the
large material hold-up, this results in contamination of the plant with reprocessed
material, which takes a long time to “flush through.” By contrast, with a very small
material hold-up in process, gas centrifuge plants are particularly suited to cam-
paigns to enrich reprocessed feedstock and with certain precautions and processes,
can switch back to enrichment of un-irradiated natural material with little “memory”
of reprocessed material contaminating the cascades.
Some examples of enrichment of reprocessed uranium are as follows:
● Until it shut down in 1982, some 16,000 tU of “magnox-depleted” reprocessed uranium
from the UK’s first generation magnox reactor fleet was fed to the UK diffusion plant at
Capenhurst to increase the 235U assay from about 0.4% to “pseudonatural” around 0.7%
assay. The pseudonatural material was then fed through URENCO centrifuge plants at
Capenhurst to form about 1650 tU of low-enriched material, which was fabricated into
fuel for the UK’s second generation advanced gas-cooled reactors fleet.● Around 1000 tU of reprocessed uranium from La Hague and reconverted to uranium hexa-
fluoride at Tricastin in France has been fed through cascades at URENCO’s Almelo plant
in the Netherlands to make reprocessed uranium fuel for Kansai in Japan, Synatom in
Belgium, and EDF in France. Minor adaptations to the enrichment plant, for example,
some remote handling and additional shielding were made to reduce the hazard from
enhanced radiation from the feedstock.● A number of utilities have sent reprocessed uranium arising from their reprocessing cam-
paigns at Sellafield in the United Kingdom and La Hague in France to Russia for proces-
sing and fabrication into fuel. In Russia, blending techniques have been deployed to
control the concentrations of 232U, 234U, and 236U in finished fuel. The Seversk facility in
Russia has been used to enrich some reprocessed uranium feedstock.
Currently, the Russian blending route is the only route available for recycle of
reprocessed uranium into fuel. There are no conversion facilities in operation
licensed to convert reprocessed uranium to hexafluoride ahead of the enrichment
step. A dedicated French reprocessed uranium conversion facility was shut down in
1998. British Nuclear Fuels (BNFL), also in 1998 abandoned a project having
partly constructed a dedicated “Line 3” reprocessed uranium conversion facility at
Springfields in the United Kingdom.
Fluorides of 232U decay chain elements, including the hard gamma emitting208Tl, are nonvolatile and may be separated out by feeding reprocessed uranium
hexafluoride to a conversion process or by any gaseous transfer of reprocessed ura-
nium hexafluoride from one cylinder to another, effectively “resetting the clock”
for the in-growth of 232U decay chain species. Fig. 12.28 shows how new radiation
in-growth increases over time, reaching equilibrium after 10 years.
Accordingly, to practically fabricate nuclear fuel from reprocessed uranium, the
conversion, enrichment, and fabrication steps need to be executed sequentially with
minimum delays, typically within 24 months from conversion to enrichment, and
within 6 months from enrichment to fabrication, to minimize radiation doses from
handling the material.
Much detail on experience handling reprocessed uranium can be found in IAEA-
TecDOC-CD-1630 (2007) available on the Internet.
350 Uranium for Nuclear Power
References
Bukharin, O., January 2004. Russian gaseous centrifuge technology and uranium enrich-
ment complex. ,www.partnershipforglobalsecurity-archive.org/Documents/bukharinrussia-
nenrichmentcomplexjan2004.pdf..
Cohen, K., 1951. The Theory of Isotope Separation as Applied to Large-scale Production of
transfer. The zircaloy fuel rods in each subassembly are arranged within two con-
centric rings around a central carrier rod. The two subassemblies are joined by a
cylinder along the plane of the center carrier rod. Each fuel assembly in the reactor
core is housed in an individual pressure tube. The total length of the fuel assembly
is 10.025 m with 6.862 m being the active region. The fuel rods contain enriched
UO2 fuel pellets and may utilize a burnable absorber (INL, 2015).
13.2.6 Uranium dioxide production
There are three conversion processes used by fuel fabricators to convert UF6 gas to
UO2 powder for the production of UO2 pellets: (1) the “dry” process, (2) the ammo-
nium diuranate (ADU) process, a wet process, and (3) the ammonium uranyl car-
bonate (AUC) process, also a wet process. The dry process results in significantly
lower quantities of liquid waste than either of the wet processes and it is the most
commonly used conversion process used today. In each of these processes, the first
step is the heating of the UF6 cylinders in an autoclave and the removal of the UF6in gaseous form. The remaining steps in the various UF6 to UO2 conversion pro-
cesses are described briefly next (NEI/ERI, 2008).
The dry process was originally developed by Siemens, now part of AREVA, and
has been licensed to several other fabricators. It consists of the hydrolyzation of the
UF6 with steam in a gas-phase reaction. This step is followed by the reduction of
the resulting uranyl fluoride (UO2F2) with hydrogen and steam in a fluidized bed
reactor to produce the UO2. The initial powder product is then calcined in a rotary
kiln with more steam and hydrogen to drive off any remaining fluoride and to dry
the powder (NEI/ERI, 2008).
In the ADU process, the UF6 gas is hydrolyzed by solution in water. Ammonia
is added to the solution producing a precipitate of ammonium diuranate (ADU).
The ADU is subsequently filtered, dried, calcined in the presence of steam and
reduced in the presence of hydrogen to produce the UO2 powder. In an earlier ver-
sion of the wet process, ammonium carbonate was used instead of ammonia, pro-
ducing AUC that was then treated in a similar manner to the ADU described
previously (NEI/ERI, 2008).
In the AUC process, gaseous UF6, carbon dioxide (CO2) and ammonia (NH3) are
combined in water resulting in a precipitate of AUC. The AUC is then processed in
a manner similar to ADU to produce the UO2 powder (NEI/ERI, 2008).
13.2.7 Fuel pellet production
Following conversion of UF6 gas into UO2 powder, the next step in the fuel fabrica-
tion process pellet production of UO2 pellets. The UO2 pellet manufacture involves
the following steps: (1) mixing of the UO2 powder with binding and lubricating
agents, (2) compaction or cold pressing, (3) sintering, and (4) precision grinding.
Binding and lubricating agents that are commonly employed in pellet manufacture
are organic compounds such as aluminum or zinc stearate and stearic acid. These
371Nuclear fuel fabrication
agents enhance the formation of pores to facilitate the increase in density during the
sintering process. The binding agents provide additional strength to the cold pressed
pellets and assist in reducing dust hazards associated with the handling of UO2
powder. The lubricating agents assist in a uniform cold pressing operation (NEI/
ERI, 2008).
Once the UO2 powder has been mixed with the binding and lubricating agents,
the next step is to cold press the resulting mixture to produce “green” pellets, which
have a density of approximately 55�60% of theoretical density. The green pellets
are then sintered in a high-temperature furnace to form a stable ceramic with the
necessary heat transfer properties. The sintering process also drives off the remain-
ing additives. Final pellet densities of about 96�97% of theoretical density are typi-
cally achieved. Once the sintering process is complete, the pellets are ground to
their final dimensions in a grinder, inspected and stored for future fuel rod loading
(NEI/ERI, 2008).
MOX fuel pellet manufacture can be accomplished via two methods: (1) dry
mixing and (2) coprecipitation. In the dry mixing process, UO2 powder and pluto-
nium oxide powder (PuO2) are ground together. Depleted UO2 powder is typically
used along with rejected UO2 pellets that have been ground into a powder. The
mixture is then cold pressed into pellets, sintered, and ground to meet final specifi-
cations in similar processes used to produce UO2 pellets. In the coprecipitation pro-
cess, a mixture of uranyl nitrate and plutonium nitrate is converted by treatment
with a base such as ammonia to form a mixture of ammonium diuranate and pluto-
nium hydroxide. After the mixture undergoes a heating process, it will form a pow-
der containing uranium dioxide and plutonium dioxide. The resulting powder is
then pressed into pellets, sintered, and ground as in the formation of UO2 pellets
(Collins et al., 2011; AREVA, 2015c).
13.2.8 Use of burnable absorbers
Burnable absorbers are neutron-absorbing materials that are commonly incorporated
in LWR fuel designs as a means of power shaping, local power-peaking control and
overall long-term reactivity control. Burnable absorbers used for long-term reactiv-
ity control, such as gadolinium (GdO2) in BWR fuel assemblies, depletes as the
fuel assembly remains in the reactor core, thereby providing greater reactivity con-
trol during initial use of the fuel (when the new fuel assembly’s reactivity is high-
est) and providing less reactivity control for the fuel assembly late in a cycle (when
the fuel assembly’s reactivity has decreased). The use of burnable absorbers varies
among vendors and fuel designs, but can be generally grouped into three classifica-
The discrete absorbers, which are rarely used today, are only employed in PWR
fuel designs and are typically incorporated in the form of an absorber-filled rod that
is inserted into a vacant rod control cluster assembly (RCCA) guide tube. Typical
absorber materials employed in these designs are Boron-10 (B10) doped Pyrex glass
or aluminum oxide-boron carbide (Al2O3-B4C) pellets. Reactivity control is
372 Uranium for Nuclear Power
achieved through variation in the absorber loading per rod, the number of absorber
rods per fuel assembly, and the total number of absorber-loaded fuel assemblies in
the core (NEI/ERI, 2008; Westinghouse, 2010).
Intimately mixed absorbers, gadolinium (GdO2) and erbium (ErO2) oxides, are
combined with the UO2 powder prior to pelletization. This burnable absorber tech-
nology has been successfully employed in both PWR and BWR fuel designs by
essentially all fuel fabricators. BWRs use exclusively gadolinia, while both gadoli-
nia and erbia are employed in PWRs. Reactivity control is achieved through varia-
tion of absorber loading per fuel rod, absorber distribution within fuel rods, number
of absorber-loaded fuel rods per assembly, and/or total number of absorber-loaded
assemblies per core.
Surface coating absorbers, such as Westinghouse’s Integral Fuel Burnable
Absorber (IFBA), are applied as thin coatings of boron compounds such as zirco-
nium diboride (ZrB2) on the surface of individual pellets in PWR fuel rods. The
coating is typically applied to specific pellets of like enrichment within a fuel rod,
or to pellets spanning a wide region of a fuel rod to achieve specific power-peaking
control (Westinghouse, 2010; NEI/ERI, 2008).
13.2.9 Fuel rod fabrication process
While the specifics of fuel rod design vary among individual fabricators and fuel
type, a typical LWR fuel rod is composed of a zirconium alloy cladding tube (such
as Zr-2, Zr-4, ZIRLO, M5, etc.), a UO2 pellet column, two end plugs and an inter-
nal plenum spring. Each fuel rod has a unique identification number to provide the
ability to trace the history of each rod manufactured. The internal plenum spring
prevents pellet movement and possible damage during handling operations. A typi-
cal fuel rod manufacturing procedure consists of the following steps: (NEI/ERI,
2008).
● The bottom end plug is inserted into cladding tube and welded in place● The UO2 pellet column is loaded into the cladding tube. It may be pushed into horizon-
tally positioned tubes or gravity loaded with the tubes tilted at an angle depending upon
the manufacturing process● The length of the pellet column is confirmed to be in accordance with manufacturing
specification through insertion of a gauge is inserted into the cladding tube● A plenum spring is inserted into the cladding tube on top of the pellet stack● The top of the cladding tube is placed in a sealed chamber for pressurization. Helium is
introduced into the fuel rod to a specified internal fuel rod pressure. If it had not been
done previously, the top end plug is inserted into the cladding tube once rod pressurization
has been achieved. The end plug is then welded to the tube. If the plug contained a hole
to achieve rod pressurization, the hole is seal welded as well● Inspections of the completed fuel rod are performed
Fuel rod weld integrity is generally verified through visual, ultrasonic, and/or
X-ray inspection techniques. Each fuel rod may be weighed for gross loading
verification and/or gamma scanned to provide verification of pellet enrichments and
orientation (Kok, 2009; NEI/ERI, 2008).
373Nuclear fuel fabrication
13.3 Current and future trends
A number of fuel cycle trends are worth noting including the continued emphasis
on fuel reliability and increases in NPP cycle length, capacity factor, and fuel
assembly burnup. The development of fast reactors in Russia, China, and India,
along with fuel for these plants is also expected to experience continued growth.
Each of these topics is discussed briefly next with references to other resources.
13.3.1 Fuel reliability
The nuclear industry continues to make progress in the reduction of fuel failures. In
the US, there was a concerted effort to meet the Institute of Nuclear Power
Operation’s (INPO) goal for zero fuel defects by 2010. This Zero by Ten Initiative
is now called Driving to Zero. The Electric Power Research Institute (EPRI), INPO,
NPP operators and fuel fabricators worldwide have worked together to develop
guidelines to address the known failure mechanisms, which include: PWR corrosion
and crud, BWR corrosion and crud, grid-to-rod fretting (GTRF), pellet-cladding
interaction (PCI), and foreign material induced failures. EPRI published an updated
Fuel Surveillance and Inspection Guidelines in 2012 and continues to review its
five fuel reliability guidelines periodically to reflect the most recent knowledge
gained through fuel surveillance and inspection programs. According to EPRI,
when the industry began the Zero by Ten Initiative in 2007, 30% of US reactors
were experiencing fuel failures. By the end of 2010, this figure was reduced to 6%
and has remained constant (EPRI, 2005, 2008, 2013a).
Even though fuel failures have been significantly reduced over the last decade,
fuel reliability remains a major industry focus. Improved debris-resistant fuel
designs and better in-plant housekeeping have contributed to reductions in debris-
related fuel failures. Vendor research and development have been focused on elimi-
nating or at least significantly reducing other types of failures while continuing to
increase fuel assembly discharge burnup. However, fuel failures resulting from
materials corrosion and hydriding, grid-rod fretting, pellet-clad interaction, etc.,
have not yet been totally overcome in either BWRs or PWRs, and axial growth and
distortion of PWR fuel assemblies and distortion of BWR channels remain major
problems for operators at some plants.
EPRI established a fuel reliability database (FRED) in 2004 to better share fuel
reliability information across the industry. FRED contains information on the fuel
types in core, the reliability of the fuel during operation, and other fuel-related
issues that affect operation. For PWRs, GTRF remains the dominant failure mecha-
nism in US reactors. For US BWRs, debris-related failures have been the dominant
failure mechanism since 2008. With most PWRs having transitioned to fuel designs
with grid fretting resistance, debris-related failures are expected to become the
dominant failure mechanism in US PWRs in the future (EPRI, 2006, 2014c).
Higher fuel duty and longer cycle lengths have increased the severity of BWR
channel distortion problems in recent years. Channels may bow, bulge or twist,
374 Uranium for Nuclear Power
altering the clearance that allows the control rod to move freely, resulting in poten-
tial safety implications due to degraded control rod performance. According to
EPRI, 17 out of 35 BWRs in the US have reported control blade interference due to
channel distortion in the last decade. Affected fuel designs include Zircaloy-2 chan-
nels manufactured by all three US fuel vendors. EPRI, US utilities, the BWR
Owners Group, fuel vendors, and INPO developed a Channel Distortion Industry
Action Plan (CDIAP) to coordinate research in this area to better understand the
mechanisms associated with channel distortion. Lead Channel Test programs have
been initiated to evaluate alternative channel materials and fuel vendors have intro-
Many of the early fast reactors, such as EBR-II in the US, utilized metallic fuel
and some fast reactor designers, such as GEH, are developing metallic fuel designs.
In metallic fuel, a metallic fuel slug is loaded into the fuel cladding and the gap
between the fuel slug and cladding is filled with sodium. The sodium acts as a ther-
mal bond until the fuel swells to meet the cladding. Fuel slugs can be the full length
of the fuel cladding tube or multiple slugs can be stacked—much in the way that
ceramic UO2 pellets are stacked in LWR fuel. The Mark-I and Mark II fuel utilized
in EBR-II was made with 95% uranium metal and a 5% fissium alloy. Fissium is a
mixture of fission products. Subsequent fuel was made from recycling the metal fis-
sion products along with recovered uranium metal. The General Electric-Hitachi
PRISM fast reactor design, which is based on the EBR-II design, would utilize
metallic fuel such as an alloy of zirconium, uranium, and plutonium (Chang, 2007;
GEH, 2015).
A new Russian fast reactor that is under development, the BREST fast neutron
reactor, will utilize lead as the primary coolant. The fuel type considered for the
first core of the BREST fast reactor is a nitride of depleted uranium mixed with plu-
tonium and minor actinides (MA). Reprocessing is limited to the removal of fission
products without separating plutonium and MA from the mix (U-Pu-MA). One of
the notable characteristics of the BREST plant and other planned fast reactors is
that a reprocessing plant is colocated with the reactor, eliminating in principle any
accident or problem due to fuel transportation (Alemberti et al., 2014; WNN,
2014b).
13.3.4 Expansion of fuel fabrication markets and new marketparticipants
The past two decades saw consolidation of fuel suppliers in the US and Western
Europe resulting in the three major Western fuel suppliers that exist today:
AREVA, GEH, and Westinghouse. The next decade may be characterized as one of
expansion of existing fabricators into new regions or the introduction of new fuel
products from new market entrants.
TVEL has historically provided almost 100% of fuel fabrication requirements to
Russian-designed VVERs in Russia, Ukraine, and Eastern Europe but it did not pro-
vide fuel to Western-designed NPPs. Recently TVEL developed a PWR fuel design
for use in Western NPPs to expand its market. In February 2012, TVEL announced
that it had signed a contract with Vattenfall Nuclear Fuel of Sweden that covers
Lead Test Assemblies (LTAs) of square 173 17 lattice TVS-Kvadrat. Conversely,
Westinghouse is the only Western fuel supplier that has produced fuel assemblies
for Russian-designed VVER NPPs, having produced VVER-1000 fuel assemblies
377Nuclear fuel fabrication
for the two Soviet-designed Temelin units in the Czech Republic until 2006. Under
an the initial 2008 contract between Westinghouse and Energoatom of Ukraine, the
first Westinghouse fuel assemblies were loaded into the South Ukraine NPP in
2010. In December 2014, Westinghouse and Ukraine’s Energoatom agreed to sig-
nificantly increase future deliveries of Westinghouse-supplied fuel assemblies to
Ukrainian NPPs through 2020.
There is also expansion of fabrication capacity in China for Western-designed
NPPs that are under construction in China. CNNC has put in place technology
transfer agreements with AREVA, TVEL and Westinghouse so that China will be
capable of fabricating fuel for its Western-designed NPPs. Kazatomprom is also
exploring joint ventures with a number of fuel manufacturers that would allow fab-
rication of fuel assemblies in Kazakhstan. Kazatomprom already exports fuel pellets
to a number of countries including China and India.
In April 2015, the Nuclear Utility Fuel Advisory Board (NUFAB), whose mem-
bers include several of the largest US nuclear operators—Dominion, Duke Energy,
Exelon, and Southern Nuclear, sent a letter to the US Nuclear Regulatory
Commission (NRC) regarding a new fuel product under development by
Lightbridge Corporation (Lightbridge). According to Lightbridge, its metallic fuel
rod design includes three components: a- central displacer of zirconium (Zr), which
serves to reduce centerline temperatures and allows for the incorporation of burn-
able poison material within the rod; a four-lobed fuel core composed of a Zr-U
alloy, and a Zr-1Nb cladding alloy that is metallurgically bonded to the fuel core.
The metallurgical bonding that takes place during the fuel rod fabrication process
results in each fuel rod being a monolithic form composed entirely of metal.
According to Lightbridge, the fuel rod design is more robust than current tubes that
utilize ceramic pellets. NUFAB members are working with Lightbridge to submit
an application to the NRC in 2017 for the use of LTA, with insertion of LTA’s into
a US PWR as early as 2020 (NUFAB, 2015; WNA, 2015e).
13.4 Sources of further information and advice
There are a wide variety of sources that provide information on fuel fabrication pro-
cesses, capacity, and fuel cycle trends. Documents from many of these sources are
cited herein and are included in the references to this chapter. Additional resources
include:
● Addition information regarding fuel fabrication can be found at the various web sites for
the fuel fabricators.● The World Nuclear Association also provides a high-level overview of the fuel fabrication
process and fuel cycle facilities in countries with nuclear power programs.● ANT International, which provides training in the area of nuclear fuel, has very informa-
tive information available on its web site, for purchase or through various seminars that
the company holds periodically.
378 Uranium for Nuclear Power
● EPRI’s Fuel Reliability Program provides information regarding fuel assembly failure
mechanism, ongoing research programs, and plant specific experience.● The proceedings of technical meetings sponsored by the International Atomic Energy
Agency, TopFuel, and other conferences also provide a wealth of information on fuel for
existing NPPs, fuel cycle innovations, and research into new fuel designs and new reactor
types.
References
Alemberti, A., Frogheri, M.L., Hermsmeyer, S., Ammirabile, L., Smirnov, V., Takahashi, M.,
Smith, C.F., Wu, Y., Hwang, I.S., 2014. Lead-cooled Fast Reactor (LFR) Risk and
Safety Assessment, White Paper, Revision 8. Gen IV Int. Forum, April 2014.
ANS, 2015. 2015 Nuclear News, Reference Special Section, Power Reactors by Type,
Worldwide. Nuclear News, March 2015.
ANT International, 2014. Peter Rudling, Advanced Nuclear Technology International, et al.
Annual Report, December 2014.
AREVA, 2010. HTP: Robust Technology for PWR Fuel Assemblies. Available from: ,http://
n/a, Not applicable; � , Very low importance; �� , Low importance; ��� , Normal importance; ���� , High importance; ����� , Critical importance.
Note: The importance rating is generic only and individual aspects of the operation and site-specific factors need to be considered.
Uranium is a naturally occurring radioactive material (NORM) that undergoes a
series of decays through the uranium decay series. Generally when mined, the ura-
nium is in equilibrium with all the decay products, which means that radiation pro-
tection from all decay products must be considered. During uranium mining, there
are three main exposure pathways to be considered: direct gamma, inhalation of
radon decay products (RDP), and inhalation of radioactivity in airborne dust.
Direct gamma exposure occurs when personnel spend significant periods of time
in proximity to large quantities of material containing uranium or a specific decay
product, radium-226. In most cases, gamma dose rate is relatively stable over time
so monitoring can easily be undertaken to identify areas with higher risk and appro-
priate corrective actions implemented. For underground mining operations gamma
exposure is the largest exposure pathway for routine practices and for high-grade
deposits can be a limiting factor for mining approaches and technology (eg, the
high-grade Macarthur River and Cigar Lake mines). Protection mechanisms for
external gamma radiation can be summarized as:
● Time: reduce the time spent in higher dose rate areas via appropriate planning, design and
equipment. For example mining from outside the immediate ore body or restricting access
from area of known high dose rate.● Distance: maximize the distance from high dose rate areas. For example situating offices
and workshops away from ore stockpiles or locating equipment controls from the active
ore face.● Shielding: placing benign material between active ore areas and the working area. For
example, shotcrete underground can be used as shielding for the walls and ceiling and
waste rock can be used as a road base in high-grade ore zones.
Inhalation of RDP are due to one of the uranium decay products, radon-222, being
gaseous. As such, it can diffuse out of the primary material and enter the air in work
areas. Radon itself is an inert, colorless and odorless gas with a half-life of 3.8 days
and as such has only a limited direct effect on the exposure to an individual (basically
it is breathed in and then breathed out giving only a short residency time in the
lungs). However, radon decays into a series of short-lived (less than 27 min half-life)
decay products. As these are particulates when breathed in, they remain within the
lung and hence give rise to the exposure pathway. RDP are generally only a signifi-
cant exposure pathway when there are significant volumes of uranium-bearing mate-
rial in an area where ventilation is restricted. Generally, it is most significant in
underground mining operations but can be significant during vessel entries, ore
reclaim tunnels, or where degassing of process pregnant liquor occurs in ISR opera-
tions. Radon is not generally an issue for the storage of the final uranium product as
by this stage in the process the parent of radon, radium, has been removed so no
radon generation. The main methods for control of radon decay product exposure are:
● Reduce radon emanation: stop radon entering the work air by blocking off mined out
zones, reduce the active radon generating surfaces in a mine, and reduce or seal ground-
water ingress.● Ventilation: provide active one pass ventilation with priority for fresh air to active work
areas and prevent recirculation of air containing radon.
390 Uranium for Nuclear Power
● Monitoring and time: Radon and RDP concentrations can change very rapidly so active
monitoring programs may be required. In areas of enhanced RDP, restriction on access or
working time can be implemented to reduce dose.
Inhalation of radioactivity in airborne dust arises when the inhaled air contains
dust particles, which include uranium and/or its decay products. Generally, only
the longer-lived decay products are significant: uranium-238, uranium-234,
thorium-230, radium-226, lead-210, and polonium-210. In the mine, all the
radionuclides can be assumed to be present but following processing disequili-
brium occurs. The tailing material is reduced in the contribution from the two
uranium radionuclides (238U and 234U). The final uranium product generally only
has those two uranium radionuclides. In most operations, inhalation of radioac-
tivity in dust only becomes the dominant radiological pathway for workers in the
final product recovery area. However, if an operation has very high dust levels
or specific dust generation activities than it can be a significant pathway in other
areas. Protection methods for this pathway are similar for general dust control
mechanisms. Specifically, dust-generating activities and processes should be
identified and controlled (ideally using engineered solutions), access to areas of
uncontrolled dust generation should be restricted, and personal protective equip-
ment should be used in areas where other methods of dust control are not
effective.
There are a number of other minor pathways of radiological exposure: inges-
tion of radioactivity, radioactivity through wounds, and adsorption of radioactiv-
ity through the skin. Normal occupational practices including personal hygiene,
designated crib area, covering of wounds, and cleanliness should ensure that
these pathways are not a significant source of exposure. Radioactivity through
wounds (injection) is only potentially significant in the event of an emergency
although generally the radiological risk is far smaller than the medical aspects of
an injury.
Radiological aspects of mining are not necessarily restricted to just uranium
mining. The mining of mineral sands are often integrally associated with expo-
sure to natural radioactive material containing thorium-232 and its decay pro-
ducts. In addition, other underground mines can provide significant exposure to
RDP particularly where ventilation is not a primary concern or where air is recir-
culated. For example, deep gold mines and some coal mines can have radio-
logical exposures far in excess of those received in modern uranium mines.
Dating back to the prenuclear era, the mining for other metals (such as silver in
areas rich in pitchblende) were associated with high radiological exposures and
resulting health impacts.
14.3.1.2 Conventional health risks
Although radiological risks are the major difference between uranium mines and
other metal mines, there remain a number of conventional (nonradiological) health
risks that require management. These risks will vary according to the mining
391Management for health, safety, environment, and community in uranium mining and processing
methodology, processing methodology, ore composition, and local site factors.
Common health risks that may need to be considered include the following:
● Physical health risks: thermal stress, vibration, noise, general airborne dust, ultraviolet
radiation● Chemical health risks: diesel particulates, silica, acids, alkalis, carcinogens, asbestos, hea-
vy metals, ammonia, acid mist, chemical irritants (though the mechanism of harm of
some of these is physical)● Biological hazards: legionnaires disease, sewerage pathogens● Ergonomic hazards: manual handling
From the health risk perspective, hearing loss from noise is perhaps the most
common health impact, followed by manual handling related impacts. With respect
to carcinogenic impacts, the highest risk is ultraviolet radiation from the sun.
Specifically associated with uranium mining, there are some more significant health
risks that are common. From the emergency perspective, the potential for a large
release of ammonia gas ranks as one of the highest risks for a health impact (ammo-
nia is a commonly used reagent for uranium precipitation). In underground mines,
diesel particulates are a common health issue and, depending on the makeup of the
ore body and surrounding rock, silica inhalation may be a concern.
It is always important to keep the risks in perspective. In absolute terms, the risk
from radiological hazards in a modern uranium mine are far smaller than the risk
from conventional health hazards. In developing an effective health regime, it is
important to keep a balanced program that addresses the more significant conven-
tional health hazards while maintaining control over the radiological hazards associ-
ated with uranium mining.
14.3.1.3 Disease prevention and health care
Although often not considered an integral part of the health aspects of a mine, the
consideration of diseases and other health impacts should be a component of the
overall HSEC program. This is not specific to uranium mining, although there may
be some areas where additional work is required due to radiological concerns.
Mining often occurs in remote areas and utilizes a workforce that may not be
indigenous to the area. This can significantly raise the risk for a range of health-
related impacts and the affect of this can result in larger consequences than all other
health (and for that matter safety) related impacts on the workforce. The hazard
may simple be due to the displacement of the workforce from their normal support
base, or may be more directly related to the operation such as the change in diet at
the mine accommodation. Issues to be considered include:
● Disease: AIDS, malaria, sexually transmitted diseases, parasites (water borne or other)● Lifestyle: obesity, mental illness, alcoholism, drug use● Location: availability of health care, climatic extremes, dangerous wildlife, distance to
hospitals and specialist treatment
The health impacts are not necessarily negative and in fact, there can be some
very significant positive health impacts. These range from the availability of good
392 Uranium for Nuclear Power
food and health care (sometimes for the first time) to positive mental health effects.
It is very common for there to be a “healthy worker” effect where the overall health
impact of a mine is positive. In these cases, workers may have significantly better
health than the general population or the population in the area where the worker
originated.
With respect to uranium mines, the disease and general health issues are identi-
cal to those for a conventional mine. The one exception is some additional concern
around radiation. Although this is not a direct effect of the radiation, it relates to
fear or social stigma associated with working with uranium and can cause second-
ary health effects. Good worker communication on radiation and potentially the use
of health screening can mitigate against these potential health impacts.
14.3.2 Current versus historical practices
The fact that in modern uranium mines radiological health risks are low is a testa-
ment to the improvement in the control of health hazards over time. When uranium
mining commenced in earnest in the 1940 and 1950s, little was known about poten-
tial health risks. As such, there were few mitigative measures in place to reduce
worker exposure and the doses received reflect this. In particular, the contribution
of radon and RDP was very high due to a lack of ventilation and close proximity to
the ore zones. Dust controls were either minimal or nonexistent and gamma moni-
toring was really only to identify the highest ore grades and not for safety purposes.
As a result of these historic practices, such as in very small-scale gouging opera-
tions, uranium mine workers received substantially higher doses than those permit-
ted today. The combination of high doses and large workforce numbers has led to a
number of epidemiology studies being undertaken on these historic mine workers.
There were proven health effects from the uranium mining and specifically an
increased risk of lung cancer primarily due to the inhalation of RDP.
In modern uranium mines, there are strict controls on radiation exposure and
annual doses are in almost all cases well below the internationally accepted limit of
20 mSv/year (100 mSv in any 5-year period with no dose above 50 mSv in any
single year). Average doses to uranium workers are typically similar to the natural
background radiation (in the range of 1�4 mSv/year) and maximum doses for most
operations are under half the limit (,10 mSv/year). At these levels, there is no
direct indication of any short or long-term health effects. In fact, the doses are less
than that experienced in a number of other professions not normally considered as
radiation related (eg, international aircrew, deep gold miners, cave guides).
For modern uranium operations, nonradiological factors have far more potential
for health impacts than radiological doses. The most common work related heath
impact is hearing loss due to noise exposure and the highest fatal health effect is
likely to be related to ultraviolet radiation from the sun (although this has signifi-
cantly reduced recently due to the compulsory use of protective clothing).
However, the impact of these nonradiological hazards is likely to be far smaller
than the impact of other health factors such as disease and lifestyle factors.
393Management for health, safety, environment, and community in uranium mining and processing
14.4 Managing safety in uranium operations
14.4.1 Introduction
The prominence of safety as the first priority for a modern mining operation is one
of the most significant historical changes within HSEC. This is irrespective of
whether the operation is targeting uranium or some other commodity although there
can be higher levels of scrutiny if it is a uranium operation.
The importance of safety is obvious from the first contact with an operation.
Before you even arrive at the site entrance, you will almost certainly see signs
advising of the importance of safety (and sometimes more general HSEC). This
comes in many different forms (zero harm, target zero, safety is the first priority,
etc.) but the clear message is that safety should be considered above all other
aspects and the aim is for there to be no safety risk. This is reinforced by the induc-
tion process, which is generally required before access is allowed. This can range
from a simple visitor induction to more workplace- and task-specific inductions for
workers. These inductions generally are required periodically and a reinforced via
regular safety meetings and toolbox talks. In the event of a safety issue, it is gener-
ally widely communicated to the workforce. Management will also have a strong
commitment to safety and this will be reflected in management targets and often
performance bonuses.
More advanced sites will often have visible measures to establish strong safety
culture among the workforce. The safety culture is designed to ensure that at all
levels of the workforce safety is actively encouraged and unsafe acts are not toler-
ated. This represents the move away from the policing type approach to a more
“community” based approach, which reflects “the safe way is the way we work
here.” From the perspective of safety, this is a far more effective approach as it is
self-fulfilling and the onus for safety is firmly with the individual.
The reason for the strong push for safety is simply that safety is good business
and moral sense. Any injury is bad for business and larger-scale incidents (such as
multiple injuries or fatalities) can be catastrophic to an operation as well as families
concerned. A modern operation is legally required to provide a safe place of work
and failure can do so can have heavy repercussions including legal action, fines,
imprisonment, and loss of license to operate. However, equally of importance is
that no one wants to be responsible for the injury or death of a workmate and this is
a very strong motivator for a safe workplace.
There are many potential levels of safety impacts that are considered by a min-
ing operation. A fundamental approach to safety is that all levels of impact should
be reported and appropriate actions taken to prevent reoccurrence. Failure in report-
ing systems are often a first indicator of poor safety culture. Terminology can
change from jurisdiction to jurisdiction but the approach is fundamentally the same.
These levels of safety impacts are:
● Incident: No injury occurs, but there was potential for an injury. These can be minor but pro-
vide an early warning of potentially more serious safety issues. The management of incidents
varies widely between operations and ranges from no action to recording and investigation.
394 Uranium for Nuclear Power
● First aid case (FAC): An injury of some form occurs and it requires only minor treatment.
This could be as simple as a paper cut or an insect bite. These are recorded both as an
indicator of potential higher risks and as a precaution for the worker in case the injury
deteriorates (it has occurred that a simple mosquito bite has become infected and given
rise to more serious safety impacts).● Medical treatment injury (MTI): A MTI is when the injury requires some form of treat-
ment beyond normal first aid. For example, if the injury requires a stitch or if some form
of medication needs to be prescribed, it is generally classified as a MTI.● Restricted work case (RWC): A RWC is when an injury occurs and because of the injury, the
worker cannot complete normal duties. The worker may be able to be reassigned to a differ-
ent task so there is no lost time. For example, a manual laborer could sprain their ankle and
not be able to do their normal task but could be assigned to the office for paperwork duties.● Lost time incident (LTI): An LTI is when a worker cannot work for a whole shift as the
result of an injury. For example, if a manual laborer seriously sprains their ankle and can-
not access site safely and hence cannot complete the following day’s shift.● Fatality: A fatality is where one or more workers lose their life due to an accident in a
workplace.● Total permanent disability (TPD): A TPD is where a worker is injured to such an extent
that they cannot work ever again as a result of their injuries.● Significant potential incidents (SPIs): An SPI is where some incident occurs or is about to
occur that could have resulted in a fatality or TPD. These can be seen as “free lessons,”
as they identify high-risk tasks where no one got hurt but the potential was there.
Common SPIs include potential for electrocution and working at heights.
MTI, RWC, and LTI are often grouped together and are considered reportable
injuries. The frequency of reportable injuries as a fraction of collective working
hours is a common key metric for safety performance. For example, the All Injury
Frequency Rate is defined as the number of reportable injuries per 20,000 working
hours by a workforce. This is considered a lagging indicator of safety performance
as it indicates previous performance. Leading indicators measure proactive safety
initiatives including incident investigations and time in field for management.
For uranium operations specifically, there is no key distinguisher from other min-
ing operations. Generally, the safety practices and indicators are identical and varia-
tions are related to the individual company’s approach to safety. The only potential
difference is that uranium operations often come under heavier government and pub-
lic scrutiny than comparable conventional mines. As such, even higher levels of safety
performance may be demanded and the consequences of safety impacts may be high-
er. Key public interest groups may use safety as a means of opposing the uranium
operation and there may be additional regulatory bodies overseeing the safety aspects.
14.4.2 Current versus historical practices
From the historic perspective, the improvement in safety in mining has been a
major success for the industry. Mining traditionally is considered a dangerous
industry to work in and this was the case in the past. However, review of modern
operations show that the safety performance of mining is now far better than most
other heavy industries and often has better safety performance than industries
395Management for health, safety, environment, and community in uranium mining and processing
considered safe (eg, safety performance of mining is far stronger than agriculture or
heavy industry and can be better than such areas as office work). Because of the
heavy scrutiny of uranium mining, safety performance is often either as good as or
better than conventional mines.
14.5 Managing environmental impacts in uraniumoperations
14.5.1 Introduction
The protection of the environment is the dominant public concern associated with
uranium mining. Uranium mining has all the conventional risks of mining with the
additional potential risk associated with radiation. Due to the long half-lives of the
radionuclides in the uranium series, the potential radiological risk is seen as very
long-term and this has greatly raised the perception of risk around uranium mining.
In terms of real environmental impacts, conventional hazards are almost always
the dominant risk. Often the biggest impact of a mine is the direct physical distur-
bance or the “footprint” of the operation. In this footprint the environment will be
directly affected either by being removed (mined), covered (stockpile, plant areas,
water/runoff storage and tailings retention structures), or modified to totally disrupt
the preexisting environment (gardens, stabilization areas). Also very significant is
the introduction of external biota (such as weeds and feral animals) which can pro-
vide both medium- and long-term impacts on an area (both on-site and along trans-
port corridors). In terms of chemical impacts, once again conventional impacts
generally dominate, including such impacts as salt deposition (from ventilation sys-
tems or road spraying) or the transport of soluble compounds in ground or surface
waters (eg, fish kills from the transport of magnesium sulfate).
Radiological impacts are generally very small and do not pose any significant
risk to the environment or surrounding human populations. However, it is of critical
importance in the minds of the public and hence is given extensive attention at ura-
nium operations. Opponents of uranium mining often quote the half-lives of the
radionuclides (238U 4.5 billion years, Th-230 75,000 years, Ra-226 1600 years) as a
reason for not allowing the mining. However, this overlooks that conventional
elements (such as arsenic, mercury, cadmium, lead, etc.) with immediate toxicity
have an infinite physical half-life so mining with these impurities have similar
long-term concerns. Uranium mining also has one unique environmental pathway,
being the generation and emanation of the inert gas radon.
As with conventional hazards, knowledge of the chemical and physical proper-
ties of each long-lived radionuclide is required to determine their transport and
impacts on the surrounding environment. Often the impacts on the environment are
incorrectly assessed due to either consideration of only the parent radionuclide, ura-
nium, without appropriate attention paid to the decay products. The behavior of all
the key radionuclides in the uranium series is complex and the following is a
396 Uranium for Nuclear Power
summary, but care must be taken as the behavior can change according to site- and
pathway-specific factors.
Uranium in equilibrium: For ore and waste rock, storage and handling the radio-
nuclides will generally be in equilibrium that simplifies the consideration of envi-
ronmental pathways. For the distribution of airborne material (fugitive dust) this
type of material often dominates. Direct exposure pathways can be assessed using
an equilibrium figure and individual determination of all radionuclides may not be
required. Secondary pathways (such as being incorporated into aquatic pathways)
will cause disequilibrium and therefore consideration of individual radionuclides
may be required.
Uranium-238 and U-234: These two isotopes of uranium are physically and
chemically identical and are almost always in equilibrium (some slight disequilib-
rium is possible due to alpha recoil of 234U, which may increase its solubility.).
The half-life of both uranium isotopes are very long (4.5 billion and 250,000
years, respectively) and as such the concentration can be measured using standard
chemical techniques such as Inductively Coupled Plasma Mass Spectrometry
(ICPMS). In the environment the solubility is very dependent on both pH and
oxidization state. As a rule, uranium is most soluble at low and high pH and
partially soluble at neutral pH.
Thorium-230: With a half-life of 75,000 years, thorium-230 dominates the half-
life of tailings deposits. Thorium is insoluble in neutral pH and is unlikely to be
significantly transported in ground or surface waters (aside from direct physical
transport). In acid tailings systems there can be crusts enhanced in Th-230 that,
when disturbed, can be aerially dispersed. Thorium-230 can be measured by
radiochemical or gamma spectrometry and the chemical measure of natural thorium
(thorium-232) may be used as a chemical proxy to determine where the thorium
reports to in a process.
Radium-226: Has a half-life of 1600 years and is generally one of the most
critical for environmental pathways due to its high solubility in near neutral pH.
It also is the parent of radon-222 and hence is the source term for radon generat-
ing structures. Finally, because the majority of the gamma radiation from the
uranium series arises from bismuth-214 (one of the RDP) the concentration of
radium-226 will dominate the gamma signature of a structure. Radium-226 is
relatively easy measured by gamma spectrometry (after sealing and decay to
ensure equilibrium with the RDP). Chemically it is similar to barium, calcium,
and magnesium.
Radon-222: Radon-222 and the subsequent inhalation of RDP dominate the
airborne pathways associated with uranium mining. Radon is an inert gas with a
half-life of 3.8 days. This time is sufficient for the radon to leave its host mineral,
diffuse through the material, and enter the general atmosphere. It can then be dis-
persed by the prevailing winds and travel significant distances from the uranium
operation. If the host material is in an area of reduced ventilation, high concentra-
tions can be experienced leading to significant radiological exposure.
Lead-210: Lead-210 has a half-life of 22 years and is most important because its
levels in the natural environment are often far higher than the other radionuclides
397Management for health, safety, environment, and community in uranium mining and processing
(by up to an order of magnitude). This is because it is the long-lived decay product
of the RDP. In the natural environment, radon escapes from the ground and decays
in the atmosphere, producing a significant natural source of lead-210. The lead-210
preferentially accumulates on near atmosphere surfaces (eg, the top layer of soil, in
foliage or specifically deposited in some plants like lichens). This natural enhance-
ment in lead-210 concentration is often misinterpreted as a uranium operation
related impact. It can be measured by beta radiochemistry or by gamma
spectrometry.
Polonium-210: Polonium-210 has a half-life of 138 days and is a pure alpha
emitter. As such, it can only be measured using radiochemistry and alpha spectrom-
etry, which means it is often missed by normal measurements. As the decay product
of lead-210 (via bismuth-210), it is also substantially naturally enhanced in environ-
mental samples. It also has a very low volatilization temperature so will be prefer-
entially emitted by operations involving heating of material. As such it often
important as a NORM in products thought of as nonradioactive (eg, coal, iron ore)
and may be important in operations where uranium is a byproduct (and part of the
processing involves heating). When measuring polonium-210 care must be taken
not to heat the sample (eg, ashing of vegetation samples), as otherwise the
polonium-210 may be volatilized out of the sample.
Tailings material: Tailings material requires a special mention from the radiolog-
ical perspective. Tailings will retain the majority of the radioactivity from the ura-
nium series (only the uranium radionuclides and their short-lived decay products
have been removed). The mobility of the radionuclides within the tailings structure
will be heavily dependent on the pH of the material and also the underlying soil
and rock may similarly change the solubility. For example, thorium-230 is mobile
in acid tailings but rapidly precipitates out of solution as the water is neutralized by
underlying soil and rock.
For both conventional and radiological hazards the common pathways of
on the characteristics of the material. Dangerous goods are defined in nine
classes internationally,1 with subdivisions under certain classes. Dangerous goods
regulations define radioactive material (which has no subdivisions)—“Class 7” as
any material containing radionuclides, where both the activity concentration and the
total activity of consignment exceeds certain defined limits. The reason for regula-
tion of radioactive material is because the radioactive decay of certain radionuclides
emitting ionizing radiation. This unique characteristic of ionizing radiation may
present a potential risk to people, property, and the environment.
Safety in transport is the responsibility of the consignor, but security is mainly
the responsibility of the states concerned.
For the safe transport of Class 7 Radioactive Material, the IAEA has published
advisory regulations since 1961. These regulations have come to be recognized
throughout the world as the uniform basis for both national and international trans-
port safety requirements in this class. Requirements based on the IAEA regulations
have been adopted in about 60 countries, as well as by the modal organizations
such as the International Maritime Organisation (IMO) and the International Civil
Aviation Organisation (ICAO), and regional transport organizations. The IAEA
has regularly issued revisions to the transport regulations to keep them up to date.
The latest set of regulations is published as Regulations for the Safe Transport of
Radioactive Material, 2012 Edition SSR-6.
The objective of the regulations is to protect people and the environment from
the effects of radiation during the transport of radioactive material.
Protection is achieved by:
● Containment of radioactive contents● Control of external radiation levels● Prevention of criticality● Prevention of damage caused by heat
The fundamental principle applied to the transport of radioactive material is that the
protection comes from the design of the package, regardless of how the material is
transported.
In 1993, the IMO introduced the Code for the Safe Carriage of Irradiated
Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks on Board
Ships (INF Code), complementing the IAEA regulations. These provisions mainly
cover ship design, construction, and equipment. The INF Code has been mandatory
since January 2001 and it introduced advanced safety features for ships carrying
mixed oxide fuel (MOX), used fuel, or vitrified high-level wastes (WNA, 2015).
15.3 Packaging of radioactive materials
The principal assurance of safety in the transport of nuclear materials is the design
of the package, which must allow for foreseeable accidents. The consignor bears
primary responsibility for this. Many different nuclear materials are transported,
1The UN Recommendations on the Transport of Dangerous Goods is the basis for most regional,
national, and international regulatory schemes.
406 Uranium for Nuclear Power
and the degree of potential hazard from these materials varies considerably.
Different package standards have been developed by the IAEA according to the
characteristics and potential hazard posed by the different types of nuclear material,
regardless of the mode of transport.
Packages used for the transport of radioactive materials are designed to retain
their integrity during the various conditions that may be encountered while they are
being transported, thus ensuring that an accident will not have any major conse-
quences. Packaging for radioactive materials includes, where appropriate, shielding
to reduce potential radiation exposures. In the case of some materials, such as fresh
uranium fuel assemblies or uranium hexafluoride, the radiation levels are negligible
and no shielding is required.
Conditions that packages are tested to withstand include fire, impact, wetting,
pressure, heat, and cold. It is important to reduce radiation doses to workers and the
public to be as-low-as-reasonably-achievable principle by adopting best practice at
the operating level. Packages of radioactive material are checked prior to shipping
and, when it is found to be necessary, cleaned to remove surface contamination.
As with other hazardous materials being transported, packages of radioactive
materials are labeled in accordance with the requirements of national and interna-
tional regulations. These labels not only indicate that the material in the package is
radioactive, by including the well-known trefoil, but also give an indication of the
radiation field in the vicinity of the package (WNA, 2015).
Because safety depends primarily on the package, the regulations set out several
performance standards in this area. They provide for five different primary
packages (Excepted, Industrial, type A, type B and type C) and set the criteria for
their design according to both the activity and the physical form of the radioactive
material they may contain (WNTI).
In the front end of the fuel cycle, industrial, type A and in some cases type B
packages can be used depending on the particular nuclear material. To prove their integ-
rity under various conditions, they are subject to a series of tests during the design phase.
Industrial packaging is suited for low-level radioactive material, such as ore,
oxide concentrate, uranium compounds, and low radioactivity wastes. The testing
condition of this package (without fissile material) includes fall resistance from a
maximum height of 1.2 m, which is variable depending on the mass of the packag-
ing being stack tested (piling packages) (AREVA).
Type A packaging is suited for an average level of radioactive material, such as
new nuclear fuels and medical radioisotopes. The testing condition of this package
(without fissile material) includes:
● Water spray test to simulate heavy rain● Fall resistance on a target2 hard surface from a maximum height of 1.2 m for packages
designed for solid materials, and 9 m for packages designed for liquid or gas materials● Penetration test using a 6-km bar released from a height of 1 m
2 In this context, IAEA Regulations for the Safe Transport of Radioactive Material, 2012 Edition SSR-6
refers to target as “The target for the drop test . . . shall be a flat, horizontal surface of such a character
that any increase in its resistance to displacement or deformation upon impact by the specimen would
not significantly increase damage to the specimen.”
407Safe and secure packaging and transport of uranium materials
Type A with fissile material:● Fire test, impact resistance at 50 km/h (drop of 9 m) on a target hard surface● Submission for accreditation from the Safety Authorities following an expert technical
inspection by the regulator, such as IRSN (Institut de Radioprotection et de Surete
Nucleaire) - French institute of radiation protection and nuclear safety in France (AREVA)
Type B packaging is suited for highly radioactive material, such as irradiated
fuels, certain radioactive sources, plutonium, and vitrified nuclear waste. There are
over 150 different kinds of type B packaging, which are tailored for varying material. The
testing condition of this package includes:● Impact resistance at 50 km/h (drop of 9 m) on a target hard surface● Head-on fall resistance from a height of 1 m● Fire resistance at 800�C for 30 min● Immersion resistance up to 200 m for the most radioactive packages● Submission for accreditation from the Safety Authorities following an expert technical
inspection by the regulator
The IAEA regulations lay down these test procedures to demonstrate compliance with
the required performance standards. The procedures, whereby the package integrity is
related to the potential hazard, are important for efficient commercial transport operations.
They also take into account the different conditions of transport characterized by the
IAEA as follows:● Conditions likely to be encountered in routine transport● Normal conditions of transport (minor mishaps)● Accident conditions
The regulations also detail marking and labelling provisions, requirements imposed
on packages during transit, and prescriptions for their maintenance (WNTI), since
most are reusable.
15.4 Security measures for transport of nuclearmaterials
In terms of security, shipments of radioactive materials must comply with relevant
physical protection requirements developed by the IAEA, as well as the security
requirements of the modal organizations such as the IMO and the ICAO. Security
concerns since 2001 have extended from theft and diversion to terrorist scenarios.
In addition, shipments comply with the security requirements of the shipping
states’ governments (WNTI). A range of protection measures are employed during
transport, ranging from the design of the package and the vehicles used as well as
security forces, access control, employee screening, satellite tracking of shipments
and coordination with local and national security authorities.
The physical protection of Class 7 materials during transport is assisted by
minimizing both the total time the material remains in transit and the number and
duration of transfers of the material, avoiding the use of regular movement
schedules, and limiting the advance knowledge of transport information including
date of departure, route and destination to designated officials having a need to
know that information (WNTI).
408 Uranium for Nuclear Power
In addition to the IAEA regulations for transport, there is a recommendation
developed by the IAEA, The Physical Protection of Nuclear Material and Nuclear
Facilities, INFCIRC 225. In this context, nuclear materials are those that carry a
potential risk of being used in a nuclear explosive device. This requires states to
take appropriate measures to ensure security and includes the physical protection
requirements for nuclear material in use, storage and during transport. Three catego-
ries of security are defined depending on the nature of the material. The nuclear
materials covered by INFCIRC 225 are principally plutonium and highly enriched235U and 233U, for which the highest security category applies. INFCIRC 225 now
extends to national as well as international transport (WNTI).
15.5 Current issues for the transport of nuclearmaterials
The difficulties associated with the transport of nuclear materials arise from several
sources, each with varying degree of occurrence and consequence.
Most transport of Class 7 materials is radioisotopes for medical and industrial
use (including some cobalt-60 sterilization sources in 4-t type B packages). But all
of it requires some understanding of the particular regulations and hence training of
people who handle the packages, so there may be cost and inconvenience to both
shippers and others handling the packages, leading to occasional denial of shipment.
Multiple layers of regulation with lack of international consistency provide disin-
centive to shippers. There are also problems of regulatory harmonization, with the
competent authority on one country not being accepted in another. Occasionally
there is de facto refusal to issue permits, and certain insurances for vessels carrying
material with more than 1% fissile content may need to be taken out by the con-
signor or consignee.
Most reports of denial of shipment relate to nonfissile materials, either type B
packages (mainly cobalt-60) or tantalum-niobium concentrates. For uranium
concentrates, the main problem is limited ports that handle them, and few
marine carriers that accept them. There has never been any accident in which
a type B transport cask containing radioactive materials has been breached or
has leaked.
Difference in regulatory structures can force delays and cost overruns, along
with the lack of harmonization pervading throughout industry. A Euratom study in
2015 identified lack of harmonization and over-regulation in transport authorization
for radioactive materials, particularly between countries, as a significant risk from a
security of supply perspective. As previously mentioned, The IAEA Regulations for
the Safe Transport of Radioactive Material, 2012 Edition SSR-6 provides guidance
for international organizations and governments to alleviate these issues.
Also, particular reactor designs may have limited supply options for fabricated
fuel, creating vulnerability to transport disruption through industrial action or
geopolitical factors.
409Safe and secure packaging and transport of uranium materials
15.6 Transport of uranium concentrates and uraniumhexafluoride
Nuclear materials have been transported since before the advent of nuclear power.
The procedures employed are designed to ensure the protection of the public and
the environment both routinely and when transport accidents occur. For the genera-
tion of a given quantity of electricity, the amount of nuclear fuel required is very
much smaller than the amount of any other fuels. Therefore, the conventional risks
and environmental impacts associated with fuel transport are greatly reduced with
nuclear power.
Transport is an integral part of the nuclear fuel cycle. There are some 430
nuclear power reactors in operation in 32 countries but uranium mining occurs in
only about 20, most of the production being from countries without nuclear power.
Furthermore, in the course of 60 years of operation by the nuclear industry, a num-
ber of specialized facilities have been developed in various locations around the
world to provide fuel cycle services. Hence, there is a need to transport nuclear fuel
cycle materials to and from these facilities.3 Indeed, most of the material used in
nuclear fuel is transported several times during its progress through the fuel cycle—
mine to conversion plant, to enrichment plant, to fuel fabrication plant, and finally
to nuclear power plant. Transport is frequently international and often over large
distances. Specialist companies organize the transport of any substantial quantities
of radioactive materials.
When radioactive materials, including nuclear materials, are transported, it is
important to ensure that radiation exposure of both those involved in the transport
of such materials and the public along transport routes is limited. Personnel directly
involved in the transport of radioactive materials are trained to take appropriate
precautions and to respond in case of an emergency.
Prior to 2001 (adoption of the INF Code), although not required by transport
regulations, the nuclear industry had already chosen to undertake some shipments
of nuclear material using dedicated, purpose-built transport vehicles or vessels.
(see later section)
Uranium ore concentrate is a material of low radioactivity and radiological
hazard. Concentrate such as U3O8 is normally transported in ordinary industrial
sealed 200-L drums in standard sea (ISO) freight containers (WNTI). No radiation
protection is required beyond having the steel drums clean and within the shipping
container. About 36 standard 200-L drums fit into a standard 6-m transport
container. They are also used for low-level wastes within countries.
Uranium concentrates pose a minor risk due to the toxicity of the powder
if it is released and ingested. In this respect, it is no different from most heavy
metal compounds such as lead oxide. Uranium hexafluoride is also of low activity
3 In the case of mining and milling, the facilities are often on the same site, reducing the requirement for
transport.
410 Uranium for Nuclear Power
and the radiological risk is minor. However, there would be a chemical hazard in
the event of a release because it produces toxic byproducts on reaction with moist
air (WNTI).
Natural uranium hexafluoride is usually shipped to enrichment plants in type
48Y cylinders, each 122 cm in diameter and holding about 12.5 tU hexafluoride
(8.4 tU). These cylinders are then used for long-term storage of depleted uranium
(DU) tails from the enrichment plant, typically at the enrichment site.
Enriched uranium hexafluoride is shipped to fuel fabricators in type 30B cylin-
ders, each holding 2.27 t UF6 (1.54 tU) (WNA, 2015). These cylinders are some
76 cm in diameter and are loaded in overpacks to enhance resistance to crashes,
fires, and immersion. The cylinder design prevents criticality, and hence chain reac-
tions. The loaded overpacks are generally transported using ISO flat rack containers
for transport to fuel fabrication plants.
Depleted UF6 or “Hex,” the residual product from the enrichment process, has
the same physical and chemical properties as natural Hex and is transported using
the same type of type 48Y cylinders.
With very few exceptions, nuclear fuel cycle materials are transported in
solid form.
15.7 Transport and packaging of plutonium
Reactor-grade plutonium is transported, following its separation in reprocessing, as
an oxide powder because this is its most stable form. Most transports are to MOX
fuel fabrication plants.
Plutonium oxide is insoluble in water and only harmful to humans if it enters the
lungs, due to alpha activity (WNA, 2015). The primary risk is due to toxicity,
except for criticality that is controlled by the package design (WNTI). Plutonium
oxide is a chemically unreactive material. Plutonium metal, though its chemical
properties are similar to those of other heavy metals like lead, is relatively insolu-
ble, and dissolves to any extent only in acid or strong carbonate solution. Because
plutonium-239 and plutonium-240 are not very radioactive, cleaning up a release of
these substances would not be expected to cause much radiological hazard (Foulke
and Weiner, 2002).
Plutonium oxide transport uses several different types of sealed packages
and each can contain several kilograms of material. Criticality is prevented
by the design of the package, limitations on the amount of material contained
within the package, and on the number of packages carried on a transport vessel.
Special physical protection measures apply to plutonium consignments. A typical
transport consists of one truck carrying one protected shipping container. The
container holds a number of packages with a total weight varying from 80 to
200 kg of plutonium oxide. A sea shipment may consist of several containers,
each of them holding between 80 and 200 kg of plutonium in sealed packages
(WNA, 2015).
411Safe and secure packaging and transport of uranium materials
In laboratories handling plutonium, a great deal of attention is paid to contain-
ment and accountability of it, and relatively few are licensed and equipped for this.
(Foulke and Weiner, 2002).
15.8 Transport of fabricated fuel—uranium and MOX
Uranium oxide fuel assemblies are manufactured at fuel fabrication plants. They
are made up of ceramic pellets formed from pressed uranium oxide that has been
sintered at a high temperature (over 1400�C). The pellets are aligned within long,
hollow, metal tubes, which in turn are arranged in the fuel assemblies, ready for
introduction into the reactor. Different types of reactors require different types of
fuel assembly, so when the fuel assemblies are transported from the fuel fabrication
facility to the nuclear power reactor, the contents of the shipment will vary with the
type of reactor receiving it.
In Western Europe, Asia, and the US, the most common means of transporting
uranium fuel assemblies is by truck. A typical truckload supplying a light water
reactor contains 6 t of fuel. In Russia and Eastern Europe, rail transport is most
often used. Intercontinental transports are mostly by sea, though occasionally trans-
port is by air. The annual operation of a 1000 MWe light water reactor requires an
average fuel load of 27 t of uranium dioxide, containing 24 t of enriched uranium,
which requires only a few trucks.
The precision-made fuel assemblies are transported in robust packages specially
constructed to protect them from damage during transport. Uranium fuel assem-
blies, typically about 4 m long, have a low radioactivity level, and radiation shield-
ing is not necessary (WNA, 2015). However, fuel assemblies do contain fissile
material and criticality is prevented by the design of the package, (including the
arrangement of the fuel assemblies within it, and limitations on the amount of mate-
rial contained within the package), and on the number of packages carried in one
shipment (WNTI, WNA, 2015).
All nuclear materials shipped by sea are in a highly stable form that is inherently
safe and resistant to the effects of outside elements. The uranium oxide and the
MOX fuel pellets are a hard, ceramic material that is so stable it can survive the high
temperatures in the core of a nuclear power plant without significant degradation.
The transport casks for MOX are massive steel structures, such as the TN 12/2,
which is made from 0.3-m thick forged steel and weighs about 100 t. The reusable
casks contain about 5 t of MOX fuel, so the vast majority of the weight is the pro-
tective casing of the cask itself. The casks measure approximately 6 m long and
2.5 m in diameter.
15.9 Ships for MOX (and used fuel or high-level wastes)
British and French nuclear companies with the Overseas Reprocessing Committee,
a consortium of 10 Japanese electric utilities, have regularly transported shipments
412 Uranium for Nuclear Power
of spent fuel from Japan to Europe for reprocessing and returned conditioned waste
and recycled fuel back again in more than 180 round-trip voyages over more than
30 years. The safety record is second to none. These ships have transported more
than 8000 t of nuclear material, and have traveled more than 7 million km, without
a single incident involving the release of radioactivity.
The three largest ships belong to a British-based company Pacific Nuclear
Transport Ltd (PNTL), a subsidiary of International Nuclear Services Ltd (INS).
They all have double hulls with impact-resistant structures between the hulls,
together with duplication and separation of all essential systems to provide high
reliability and also survivability in the event of an accident. Twin engines operate
independently. Each ship can carry up to 20 or 24 transport casks.
The three PNTL vessels now in service, Pacific Heron, Pacific Egret, and Pacific
Grebe, were launched in Japan in 2008, 2010, and 2010 respectively. They are 4916 t
deadweight and 104 m long. Pacific Grebe carries mainly wastes, the other two
mainly MOX fuel. Earlier ships in the PNTL fleet mainly carried Japanese used fuel
to Europe for reprocessing. The PNTL fleet has successfully completed more than 180
shipments with more than 2000 casks over some 40 years, covering about 10 million
km, without any incident resulting in release of radioactivity (WNA, 2015).
The ships used to transport MOX fuel are among the world’s safest. The fleet is
certified to INF3 (Irradiated Nuclear Fuel class 3)—the highest safety category of
the IMO for nuclear voyages. The vessels have been designed and built specifically
to carry these nuclear materials as well as high-level nuclear wastes, and they
employ a range of safety features far in excess of those found on conventional
cargo vessels (Foulke and Weiner, 2002).
For the radioactive material in a large type B package in sea transit to become
exposed, the ship’s hold (inside double hulls, if INF3 vessels) would need to rup-
ture, the 25 cm-thick steel cask would need to rupture, and the stainless steel flask
or the fuel rods would need to be broken open. Either borosilicate glass (for repro-
cessed wastes) or ceramic fuel material would then be exposed, but in either case,
these materials are very insoluble. However, these materials are outside the scope
of this book.
The INF3 transport ships are designed to withstand a side-on collision with a
large oil tanker. If the ship did sink, the casks will remain sound for many years
and could be recovered because instrumentation, including location beacons, would
activate and monitor the casks (WNA, 2015).
15.10 Sources of further information
Dangerous Goods International, 2014, The 9 classes of dangerous goods http://
www.dgiglobal.com/classes#rad
University of Ottawa, 2011, Transportation of Dangerous Goods Class 7
The objectives of mine site remediation and reclamation include the following:
� To remove hazards due to tailings, waste rock and contaminated water to
acceptable values and standards
� To protect the environment in the long term
� To address any risks, possibly only perceived, of concern to the local population
The following sections of this chapter briefly discuss the different uranium
mining methods and milling processes (which are described in more detail in
Chapters 6 and 7), the radiological implications in regard to site remediation and
reclamation and the methods and tools used to carry out site rehabilitation. They
do not examine, in detail, site legacy and cases of poor past practice (see
Fig. 16.1), but references are provided to enable the reader to further investigate
that subject. Because the remediation of legacy sites has been extensively
reviewed by the International Atomic Energy Agency (IAEA) and other organiza-
tions (OECD, 2014), this chapter on remediation focuses on best practice in the
remediation of proposed new uranium mining sites. There is an opportunity to
remediate legacy sites, if these existing sites are revisited for further development
for uranium or other potentially valuable commodities. Examples of leading prac-
tice in uranium mine site rehabilitation, as reported in the literature, are available
for a variety of specific site scenarios (mine type, milling processes used, waste
storage facility design) to assist with planning going forward.
Figure 16.1 Legacy uranium mining site in Central Asia.
416 Uranium for Nuclear Power
The IAEA, OECD/NEA, International Commission for Radiation Protection
(ICRP), and the WNA, as well as national regulatory bodies and corporations, pro-
vide invaluable forums which gather data and provide tools for regulators and
operators to best manage the mining and milling of uranium to minimize any detri-
mental impact of the industry (Woods, 2013), particularly in regards to the potential
radiological impact. Industry has held regular meetings with the IAEA, as well as
other international bodies, to provide up-to-date information on the subject of ura-
nium mine site rehabilitation. One notable example is the UMREG/IAEA meeting
based historically on the substantial recent experience gained in Germany on legacy
uranium mine site remediation.
Recent requirements for stewardship in sourcing uranium for nuclear power
plants has advanced the need to adopt leading practices in mine and milling site
rehabilitation even where national regulatory requirements are less restrictive. It is
expected that best practice principles will continue to develop into the future and
that regulators and operators will be knowledgeable and will adapt to such develop-
ments (IAEA, 2010), as a requirement for sustainable development. Best practice
can reduce the overall cost of a project particularly those costs associated with post-
mine and postmill closure.
Research in the uranium mining and milling industry continues for the place-
ment, containment, and isolation of radioactivity in mine wastes to provide effec-
tive long-term containment without significant financial, health, or environmental
liabilities placed on future generations.
Although the radiological impact of uranium mine wastes may specifically be
small, the overall effects in combination with other toxic contaminants, such as
acidity, alkalinity, heavy metals, turbidity, salinity, etc. increase the need to address
the combined environmental impact.
There are risks associated with long-term waste disposal in the light of the
requirement for the longer-term stability of tailings and for minimal environmental
impact. Many of the risks and methods of addressing these risks have been identi-
fied in previous studies. Further developments are expected to continue to address
these risks in the future, for example, the possibility for the removal of other radio-
nuclides or radioactivity-containing minerals from tailings (which has proven
difficult to achieve in past research for specific sites) to minimize the radioactivity
present, in the long term, in large quantities of tailings deposited on the surface.
Postclosure management of mine sites is a complex subject of worldwide reg-
ulatory interest, as shown in recent Canadian guidelines (Cunningham et al.,
2015). The Canadian uranium mining industry is a world leader in the adoption
of best practice and has suggested a procedure, shown in Fig. 16.2, to assist in
planning a uranium mine and mill site. This is an example of a procedure from
mine and mill site operation up until the return of the site to the authorities or
landowner, with a system of funding to achieve the necessary rehabilitation and
monitoring.
417Uranium mine and mill remediation and reclamation
2. Operator decides to end operations – Submits final
closure plan to SE and CNSC (U only) for environmental
assesment review and approval.
6. Operator applies for release – Submits application
for “Release from decommissioning and reclamation” to
SE for review and approval and application for release from
licence to CNSC (U only).
7. SE and CNSC (U only) conduct detailed review
of application
9. SE issues “Release from
Decommissioning and Reclamation”and
CNSC issues license exemption (U only)
10. Operator applies and is issued a
release from the Provincial Surface Lease
Agreement – The operator (or site holder) is
solely responsible for the site until released.
11. Operator applies for entry of site into
the Institutional Control Registry – The
Province of Saskatchewan accepts on the
condition that items 9 and 10 have been
approved. In practice, approvals issued in
items 9 to 11 would take effect the same day.
1. SE – Saslatchewan ministry of environment
2. CEAA – Canadian environmental assessment agency
3. CNSC – Canadian nuclear safety commission
3. SE and CNSC (U only) approve closure plan –
Following approval of a closure plan, license, and approval
to decommission are issued.
4. Operator implements final closure plan
Additional rehabilitation
and monitoring
8. SE and/ or CNSC (U only) refuse to
issue a release and require the site
owner to complete further rehabilitation
and/ or continue the trasition phase
monitoring program
Note: Dependent on the nature and potential
environmental impact of a development proposal,
as environmental assessment may or may not be
required. The determination is based on the
federal Canadian Environmental Assessment Act,
the Canadian Nuclear Safety and Control Act and
the provincial Environment Assesment Act and
various regulations. The Canadian Environmental
Assesment Agency facilitates the federal
assessment process rather than issue specific
approval.
5. Decommissioning and reclamation completed–
Site owner implements transition phase monitoring.
1. Site operating – SE1, CEAA2 and CNSC3(U only) have
approved conceptual closure plan in initial environmental
assesment review and approval.
Figure 16.2 Progression of a mine/mill/waste management site closure and custodial
transfer.
Source: Cunningham et al. (2015).
418 Uranium for Nuclear Power
16.2 Uranium mine and milling sites
Mine types and milling processes are reviewed in Chapters 6 and 7. These methods
are briefly revisited in this section as they affect uranium mine and milling site
rehabilitation. All aspects of the mining, processing, and waste management at a
uranium mine site must be planned and monitored in accordance with best practice.
Elevated concentrations of radioactivity contained in site wastes must be managed
in accordance with international standards and local regulations pertaining to the
management of radioactivity.
The process of rehabilitation of a site begins with an environmental impact
assessment and background, baseline study of the project before mining begins.
The ultimate impact of a project is assessed during rehabilitation but the impact
must be monitored continuously during the project from commencement of the
project to mitigate adverse long-term effects and to alter the plan during operation,
if the circumstances change. The environmental impact assessment is carried out to
identify potential impacts of mining and milling and to put in place measures to
address any identified potential risks.
Risk assessment, good management (including quality control (QC) and quality
assurance (QA) systems) and sensitivity analysis of planned changes are essential
components of the overall planning and management processes, from exploration to
long-term, postclosure stewardship.
16.2.1 Introduction to mining and milling methods
Uranium ores are generally very large hard rock type deposits such as nonconfor-
mity, calcrete, or carbonaceous requiring mining, crushing and grinding, or porous
sandstone deposits amenable to heap leaching, see Fig. 16.3, or in situ leaching
(ISL). The orebody determines the mining method: surface open cut (see Fig. 16.4),
underground, or in situ. The nature of the orebody, the ore grade (wt% U) and the
mineralogy determine the ore mineral processing (beneficiation) and extraction
methods used. These all, in turn, are determinants of the methods used for waste
disposal and options for site rehabilitation. Closure will depend on stakeholder
requirements depending on the location.
“Designing for closure” involves the development of closure objectives as a
component of the overall project design. Sustainable remediation and closure objec-
tives are site specific (IAEA, 2010) and involve issues including:
● Final or sequential land use● Human health and safety● Social impacts● Ecosystem impacts● Regulatory requirements● Cost optimization
419Uranium mine and mill remediation and reclamation
Although the evolution of standards and regulations might be regarded as poten-
tially moving the goalposts and a source of risk, this must be transparently
addressed in discussion with the regulators and other stakeholders. Regulatory
requirements must be expected to evolve with increased knowledge and should not
necessarily become more restrictive.
Liner
system
Heap
Acid drip
Acid
recirculation
Uranium recovery
Slope
Collection
basin
Processing
plantConcentra
ted
Extracted
Stripped
Drying
Driedyellow-cake
Figure 16.3 The Heap Leach Recovery Process (US NRC).
Figure 16.4 Open cut former uranium mine in Australia.
420 Uranium for Nuclear Power
The additional potential risks from radioactivity must be addressed in rehabilitation.
Pathways to exposure to radioactivity at uranium mining and milling sites include:
● Ingestion (contaminated plants and food and/or water)● Direct exposure to gamma radiation● Inhalation of dust, radon and radon progeny
With these potential pathways to exposure to radioactivity in mind, design
options and optimization for successful decommissioning and remediation must
iteratively consider and then reconsider:
● Mining options available● Milling/processing options available● Waste stream storage and disposal options● Water management and treatment options● Monitoring requirements● Remediation options● Mine and mill location options
The process to approve exploration and mining may specify the required site
rehabilitation as indicated in the South Australian Royal Commission into the
nuclear fuel cycle (South Australian government, 2015). The mining and extraction
methods employed will determine the radionuclides present in wastes (gaseous,
solid, and liquid) and the potential for their release into the environment. Baseline
data collection (of local soils, waters, and plants) is carried out during the explora-
tion drilling program to provide a reference point for future impact assessment.
16.2.2 Mining methods
Mining methods used include open cut, underground and ISL methods. Each of
these has very different potential environmental impact advantages and disadvan-
tages with regard to site remediation requirements. The deposition of solid wastes
into a surface or underground waste management facility (WMF) is of fundamental
importance in regard to the stability and permeability of the final facility and the
longer-term environmental impact of its remediation.
The deposition of waste (tailings and other process wastes, contaminated equip-
ment and mine waste) either onto a surface or into an underground facility must
produce a waste structure with suitable characteristics for the long term. Backfill or
cemented backfill of tailings as a paste into the WMF are techniques used to dis-
pose of tailings and to stabilize the mine to improve mining extraction. These depo-
sition methods can produce a structure with reduced permeability to surface and/or
ground waters and to erosion and therefore the likelihood of impurity (including
radioactivity) dispersion into the environment. The surface disposal of slimes con-
taining higher radionuclide concentrations, with greater propensity to dispersion as
dust and/or to erosion has led to the development of “whole-of-tailings” in paste
techniques being used as backfill underground or into pits. These deposition
421Uranium mine and mill remediation and reclamation
methods are likely to be accompanied by barriers, on the surface or underground, to
limit gaseous (radon) and solution migration through the waste.
16.2.3 Extraction processes
Uranium extraction processes include tank leaching, heap leaching, and ISL. These
processes determine the nature of the wastes arising for disposal (solid and liquid),
the particle size distribution (PSD) of the waste solids, the mineralogy and chemis-
try of the waste, the hydrological properties of the wastes and the physical proper-
ties of the waste. Pollutants from processing may be released (ARPANSA, 2014):
● As solids● Into the atmosphere (as gases and dusts)● Into the aquatic environment (surface waters)● Into groundwater
The properties of the process wastes determine the rehabilitation risks, for exam-
ple, the amenability of the site to decontamination and/or isolation during rehabilita-
tion. The classification of wastes, see Fig. 16.5 and appropriate handling and disposal
methods are discussed extensively in the literature (Williams, 1998).
The impact of extraction processes, as well as determining the solids tailings dis-
posal methods used, will also determine the nature of short and long-term
Critical bearing pressure (kPa)
0.01 0.1
0.6 6.0
0.30.03
Centrifugal pumps
Positive displacement pumps
“Raining”
? ? ? ?
Beaching
Intensive
improvements Conveying
Foot traffic
D3
Fluidtailings
Watercapped
Soft groundstrategies
Normalterrestrial
reclamation
Very softtailings
Softtailings
Hardtailings
D11Dozers
Pickup
Haul trucks
End-dumping
Liquid
limit
“Mobilized” undrained shear strength (kPa)
1 10
60
3.0
600
100 1000
Critical cap thickness (m)
Pumping limits
Capping method limits:
Trafficability limits
Terminology
Reclamation methods
Note: “Limits” refers to lower bound operational strength.
-Hydraulic
-Dry land
Figure 16.5 Proposed classification of tailings for remediation.
Source: From Jakubick et al. (2003).
422 Uranium for Nuclear Power
wastewater treatment requirements. They also determine the longer-term monitoring
and assessment required, as well as the likelihood of the need for appropriate dis-
posal of plant and equipment during and after plant operation.
Time, distance, and the shielding provided by the waste placement method used
and of covers determine the radiological dose received in the environment, as evi-
denced by the resilience of local bioavailability, bioaccumulation potential and the
likely impacted species. Assessment of the impact from exposure to radiation
(Larsson, 2004) is required for the short and long term.
Uranium extraction processes that facilitate mine closure must be adopted in the
future.
16.3 Site remediation and reclamation
The expectation of society is that good practice, based on good science, will be
used for site rehabilitation in adherence with regulations. In the US, the nuclear reg-
ulator (US Nuclear Regulatory Commission, 1997) focuses on the following for the
regulation of uranium mine sites, and its regulations are frequently followed in
other countries:
● Reducing radon emanation● Preventing the spread of contamination through erosion● Reducing contamination by seepage● A thorough assessment of risk to the public and the environment● Risk of groundwater contamination
The US Nuclear Regulatory Commission (NRC) has agreements in place, which
authorize most of the individual states to regulate the sources of radiation that the
NRC does not. This generally includes all naturally occurring radioactive materials
(such as radium and radon) within their borders—see http://www.nrc.gov/about-nrc/
radiation/protects-you/reg-matls.html.
The choice between alternative rehabilitation methods to minimize the environ-
mental impact of a uranium mine site is very site specific, not only because of the
differences in mining and milling methods used as discussed previously, but also
because of the climate differences and the nature of the surrounding environment.
Other local factors that must be considered in developing the necessary plans
(Williams, 1998) include local regulatory requirements and local stakeholder
requirements. Local water pathways are also critical in the evaluation of remedia-
tion options.
Guidelines have been developed by the geotechnical professions in many coun-
tries (IAEA, 2004) for the design, construction, operation, and rehabilitation of tail-
ings dams for the mining industry. Efficient and cost-effective technologies for mine
site rehabilitation of these areas (Saskatchewan Research Centre CLEANS project
and the German Wismut (UMREG, 2013) rehabilitation projects) for use by future
generations are required. Long-term rehabilitation measures are necessary with
agreed objectives between stakeholders based on sound science and good
423Uranium mine and mill remediation and reclamation
management. Uranium Mill Tailings Remediation UMTRA in the USA (G. J. Rael
Reference 18 in (IAEA, 2004)) involved extensive US government funding of designs
for storage or disposal in robust waste management facilities for up to 1000 years.
More recent research focused on the reduction of water infiltration through the
facilities.
The design criteria for remediation are determined by:
● Remediation objectives and aesthetics, as required by the stakeholders● Geotechnical, radiological, geochemical, ecological and hydrological factors as deter-
mined by the local site conditions
Implementation of best practice minimizes the potential for adverse environmen-
tal, social, and economic impacts. Best practice begins at the project conceptual
stage with natural background baseline data collection and continues through to site
remediation, closure, and site stewardship.
Other site-specific conditions such as local climate (wind, rain) and terrain (trop-
ical, desert, etc.) and proximity to populations also determine the appropriate reha-
bilitation method options. Site decommissioning, remediation and land reuse are
integral to the overall project plan and will require the application of principles
such as the as-low-as-reasonably-achievable (ALARA) principle to meet environ-
ment, social, economic and governance objectives. Baseline data collection for local
soil, plants, and water sources (IAEA, 2010) and the development of stakeholder
relationships will establish the basis for future environmental and social impact
assessment.
Remediation strategies must, to the maximum extent possible, produce a final
landform structure that is maintenance free with minimal need for ongoing manage-
ment or for future intervention. The commencement of remediation as early as pos-
sible during operations is the most cost-effective practice, as rehabilitation
addressed as a legacy has proven expensive and uncertain to happen.
Compliance monitoring of parameters, as agreed with stakeholders, during and
after decommissioning and monitoring to ensure closure objectives are met are part
of the continual improvement process, particularly if the site facilities are classified
radioactive and are regulated as such.
16.4 Risks addressed in rehabilitation
The presence of radioactivity at a uranium mine site increases public sensitivity and
potentially increases the risk to the operation compared with that for the broader
mining industry. Several studies have indicated a potential increased risk (Williams,
1998) arising from doses to humans and biota from uranium mill and mine wastes.
Early planning will help identify those potential risks and provide remedies to
address them as early as possible, which will also reduce the associated costs. This
approach, together with gaining consensus with stakeholders as soon as possible,
limits expensive data collection to only that necessary. The following sections
424 Uranium for Nuclear Power
discuss some of the risks faced by future uranium mine site operations in regard to
site rehabilitation.
16.4.1 Regulatory risk
Best practice radiation protection is expected in the regulations (ARPANSA, 2005),
however, changes in regulations are a significant risk to a project. Radiation protec-
tion philosophy and standards are continuously evolving. Regulatory reviews and
changes can place unforeseeable risks on a project and can have a significant
impact on site closure and release of the site and on the potential land end-use
criteria (IAEA, 2004).
Codes and standards are available which provide the regulatory framework in
which to manage and protect the environment. This includes the requirement for a
radiation management plan (RMP) and a radioactive waste management plan using
best practicable technology. This includes justification of practices and obtaining
authorization. The regulator, which provides approvals, authorizations and reporting
requirements, considers the potential sources of exposure to radiation.
Regulation changes will only be effective if they are realistic.
16.4.2 Effect of radiation on nonhuman biota
Previously, it was considered that in protecting humans from exposure to radiation,
other organisms would also be protected (UNSCEAR, 2008), that is, that higher
order organisms (such as humans) will experience the effects of radiation at lower
doses than would lower order organisms. In recent years, the ICRP has suggested
the need to also assess the impact of radioactivity on the biota in addition to its
effect on people. Several reviews have identified the shortage of data on this subject
for specific species and the limited modeling tools with which to make the assess-
ments (Twining, 2012). A graded approach to dose limits for this impact is recom-
mended by some regulators with “screening levels” of dose above background
(ARPANSA, 2014).
16.4.3 Solids waste disposal
Waste management during operation and the proposed disposal methods are
included in the justification of the practice. Waste rock and mill tailings may
require both active and passive controls to limit the longer-term leaching and
migration of pollutants arising from these wastes. An action plan is required to
address the risk of failure of the installed systems to contain pollutants, as measured
by site monitoring and affect assessments.
16.4.4 Acid mine drainage
Acid mine drainage (AMD) is an important aspect of uranium mine/milling site
rehabilitation (Merkel et al., 2002), as it is in other mining industries, where
425Uranium mine and mill remediation and reclamation
contained sulfides can oxidize and produce acidic solution. AMD results in an
increased potential to mobilize impurities, including radionuclides of uranium,
radium, thorium, and lead. Ground and surface water treatment may be required
continuously or intermittently in the short, medium, or long term depending on the
concentrations of these elements, which are more likely to create concern as heavy
toxic metals, rather than as being radiologically toxic.
16.4.5 Waste facility design
Above ground, below ground, and subaqueous disposal options have been used in
the past for uranium mine waste disposal. The deposition of tailings above the
ground may quarantine the land or limit its long-term use. These types of waste
facility can also be more susceptible to catastrophic failure—see Fig. 16.6.
Deposition may also make use of mined out pits. Designing a waste facility for the
long term is challenging. The ability to validate models predicting long-term behav-
ior is the subject of ongoing research and development.
Although underground placement of wastes has many obvious advantages over
above ground deposition (more aesthetic and the wastes is less prone to erosion and
wind dispersion), the contamination of groundwater, particularly if the wastes are
Tailings beach
Tailings beach
Tailings beach
Decant pond
Gully
Side-hill
Decant pond
Paddock
or
ring-dyke
Embankment
Embankment
Embankment
Decant pond
Figure 16.6 Types of above ground impoundment.
Source: From Williams (1998).
426 Uranium for Nuclear Power
below the water table, requires special consideration and the effects will be site spe-
cific. Not all the tailings may fit back into the available underground mined space.
Maximized settled sludge densities increases the ability of an impoundment to
accommodate wastes as well as reduces the permeability of above or below ground
deposition (groundwater ingress is reduced). Sealing and injection of grouts are
techniques used to help produce barriers to flow into and out of the structure.
Cemented paste backfill or high-density backfilling using the whole-of-tailings PSD
assist the facility to also accommodate troublesome slimes. Diversion of groundwa-
ter around waste masses by use of appropriate deposition and barrier techniques is
also practiced.
A sound plan with a specific waste facility design, setting clear objectives and
goals, a flexible project scope, with schedules, budget, project organization, setting
roles, responsibilities, procedures and milestones will help to ensure that these risks
are addressed using best current practice.
16.5 Tools used in rehabilitation
Concerns for the natural environment and land reuse at a uranium mining and mill-
ing site include:
● The risk of environmental degradation● Contamination of water, plants and soil● Reduced ecosystem viability and biodiversity● Aesthetics● Public amenity● Access to land● Quarantining of land for future beneficial land use
Uranium mill tailings are of particular environmental concern because:
● They contain most of the radioactivity originally in the ore● The contained radioactivity is long lived● They may also contain a range of biotoxic heavy metals and other compounds● They may contain sulfidic minerals and be prone to produce AMD● The particle size ranges from slime to approximately 500 µm making them susceptible to
leaching, erosion, burrowing by animals and to having poor consolidation properties● They often have large surfaces in contact with the elements thereby having high radon
emanation, are prone to dust generation, interaction with surface water systems, etc. under
adverse weather conditions● They occupy large areas of relatively shallow depth, which sterilize otherwise valuable
land from other future uses
Waste may also be in the form of low-grade ore or mine waste, as heap leached
material or of other wastes from processing (scales, gypsum precipitates, etc.). The
primary objectives of the isolation and the stabilization of mining and milling
wastes is achievable using a systematic approach to decommissioning and rehabili-
tation (IAEA, 2004).
427Uranium mine and mill remediation and reclamation
Design for the long term and impact assessment tools are discussed in the
following.
16.5.1 Storage facility design for closure
The design criteria for a future uranium mine site are based on the characterization
(mineralogy, chemistry, physical) of the tailings and waste rock, such as hydraulic
conductivity, drainage, consolidation, etc. properties towards achieving an agreed
plan for closure of the site. These criteria must be available at the environmental
impact statement stage of a project. Factors taken into account in design include:
● The expected life of the facility to meet regulatory integrity requirements● Cover design to prevent dispersion from waste● A radon barrier to limit the dose from this source to agreed values● Surface water runoff design to minimize erosion and permeation through the waste● Erosion prevention measures● Moist/wet tailings deposition options including under water (subaerial) disposal● Geotechnical considerations● Nonradiological as well as radiological contaminants● Drainage from the waste● Infiltration of water into the facility● Spread of contamination by wind● Control of runoff from contaminated areas● Monitoring and periodic maintenance● A monitoring plan and dewatering/monitoring wells
Many older designs such as “valley containment” and “marine disposal” are no
longer considered best practice, depending on local conditions. Any design of a
dam that allows uncontrolled discharge of tailings liquor is also not acceptable.
The failure of past waste storage facility designs has led to new dam designs,
such as those using permeable reactive barriers and liners to sequester mobile con-
taminants. Designs have also used wicks to dewater the structure prior to covering
to assist consolidation and long-term stability.
16.5.2 Modeling and model validation by monitoring
Models are used extensively to include hydrological, geochemical and geotechnical
aspects into alternative site rehabilitation options for the prediction of contaminant
transport (Merkel et al., 2002; Merkel and Hasche-Berger, 2006). Validation of the
models provides confidence in the prediction of the long-term behavior at the site.
The various models available to study the behavior of radioactivity (and their lim-
itations) are closely compared in IAEA documentation (IAEA, 2007) and can be
used to compare waste facility design options for the prediction of long-term
impacts and to assess impacts at different stages of a project. Much of this is based
on data available on the web such as that from the US, for example, the RESDRAD
codes, which are available at https://web.evs.anl.gov/resrad/.
A range of engineering measures are taken to address the potential adverse impacts
of a uranium mine and mill site using containment, covers, water treatment, etc. to
achieve the isolation of waste material and contaminants from the environment dur-
ing both project operation and rehabilitation phases. The physical properties of the
material are altered to enhance the strength of specific parts of the structure, while
some parts of the facility are designed for water and slime retention during opera-
tion, which determines the facility ultimately requiring rehabilitation. Maximum
physical stability of an impoundment reduces the need and cost for subsequent
dewatering during both decommissioning and rehabilitation.
16.5.3.1 Covers
If above ground mining and mill waste disposal are unavoidable (completely or
partly), the use of appropriately designed multicomponent/layered covers over the
facility is best practice. Near surface disposal in the vicinity of a mine and mill can
pose serious engineering challenges depending on the geomechanical and physical
and chemical characteristics of the wastes. Covers are designed to:
● Minimize radon and dust dispersal● Shield the environment from gamma radiation● Minimize water and oxygen infiltration● Control erosion● Provide an acceptable landscape in the longer term
The integrity and time guarantee on the performance of covers for 200�1000
years has been studied by the US EPA, though the contained radioactivity takes lon-
ger to decay. Covers may need to be stable for a very long time depending on the
concentration of radioactivity in the wastes and cover design. Care and maintenance
is expensive but long-term stewardship is expected. Covers should exploit the
expected ecological changes in and on the cover rather than have a cover, which is
purely a physical barrier vulnerable to erosion and other forms of deterioration. An
appropriate cover design harmonizes natural forces and incorporates vegetation and
landscape evolution. A cover’s performance should ideally improve over a 1000-
year period. Designs are site specific based on local data. Cover designs for differ-
ent climatic conditions have been studied (Benson, 2002).
The thickness and nature (particle size, moisture content, etc.) of covers to
abate radon flux to agreed satisfactory values can be calculated using the cover
material characteristics, the predicted prevailing weather conditions, etc. Covers
incorporating low concentration uranium mining wastes were examined in a
French mine site study and the results compared with those obtained using model-
ing. The calculation included cover material characterization, hydrology, and
meteorology (Ferry et al., 2002).
Australian radon flux measurements examined the effect of seasonal variation
(Bollhofer and Doering, 2015) on radon flux. The application of models for multi-
layered covers to achieve a defined threshold radon emanation has also been
429Uranium mine and mill remediation and reclamation
reported (Dinis et al., 2011; IAEA, 2013). Work is largely based on earlier guide-
lines (US NRC, 1989) and research.
16.5.3.2 Barriers
Secure containment of pollutants in a waste facility is an essential objective and
barriers assist in achieving this. This requires an understanding of the natural bar-
riers available as well as designs for engineered barriers (Metcalfe and Rochelle,
1999).
Permeable structures surrounding a waste facility allow groundwater to flow
around the waste and minimize the leaching of contaminants from the waste.
16.5.3.3 Dewatering
Excess water in a waste facility may be extracted and treated during operation to
reduce the need for more expensive postdecommissioning dewatering prior to cov-
ering and reclamation. The disadvantages and advantages of various dewatering
methods are described in IAEA publications (IAEA, 2004) and can have a signifi-
cant impact on the time and cost for subsequent rehabilitation.
The final stage of site rehabilitation involves reclamation of the area for reuse. It
involves top dressing of soil, profiling, and revegetation. This may be done once
the objectives for pollution control have been achieved. It should make use of expe-
rience available from, for example, municipal landfill design and reclamation, but
subsequent measures, such as the nature of replanting and construction on the
reclaimed site, must not interfere with the controls previously placed to minimize
radiation dose and the release of other pollutants.
16.5.4 Water treatment
Residual contaminated process water from the mine and mill and contaminated
groundwater may need treatment during and after plant operation before release to
the environment. The presence of excess carbonate containing neutralizing agents
used to neutralize free acidity in contaminated water may increase the dissolution
of uranium and its mobilization in the longer term. This has placed attention on
plant neutralization process procedures and conditions and the examination of other
(including lower cost) alternative neutralizing reagents. The use of processes such
as high-density sludge (Merkel et al., 2002; Topp et al., 2003) and reverse osmosis
are being used to treat excess process water before its discharge to the environment.
Radium removal using barium chloride and ferric sulfate is also standard practice,
when required.
The site water balance is closely monitored during decommissioning and
● Natural attenuation with monitoring● In situ remediation
430 Uranium for Nuclear Power
● Extraction of contaminated groundwater and treatment● A combination of these processes
The objective is to eliminate seepage from the ground being remediated. Natural
attenuation measures may not meet the clean-up schedule criteria. Collection and
treatment may be necessary but ultimately (in the longer term) the aim is to have
this done passively. However, initial extraction and treatment may speed up the
remediation process. The costly need to treat AMD (IAEA, 2004) should be
minimized by suitable facility design in regard to covers and barriers, surface water
runoff and groundwater flow.
16.5.5 Communication and administration
An examination of the models used in the assessment of the environmental impact
from radioactivity (IAEA, 2007) includes the statement that there should be a
“strong emphasis on good communication between all stakeholders involved in a
particular evaluation.” Communication between all stakeholders in all aspects of
the rehabilitation of a facility is essential. This must include a discussion of regula-
tory, legal, environmental, and economic issues and risks, and public concerns. The
overall objective and the progress of a rehabilitation project must be clearly com-
municated. This ensures that all stakeholders understand and trust the process as
being transparent.
Good communication and cooperation is also critical between the different
groups of scientists involved, to ensure that the requirements of the different groups
are clearly understood by everyone involved in the project.
Poor communication with stakeholders may result in the stakeholders not
“accepting” the project.
16.5.6 Training
Training of staff is essential for the safety and effectiveness of any mining project.
It is essential also for monitoring, modeling, impact assessment, reporting, etc., for
closure.
Modelers, for example, need to be well trained in the use and application of the
available models, and particularly in the interpretation of the results produced by
the models. Poor training can lead to misinterpretation of model predictions.
There is the additional training requirement when processing and managing
wastes from uranium mining and milling due to the presence of elevated concentra-
tions of naturally occurring radioactivity. Specialist knowledge and training is
required over a broad range of subjects to address this. They include professionals
such as radio-analytical chemists, metallurgists and chemists with an understanding
of the deportment of radionuclides in processing, modelers, and biologists with spe-
cialist understanding of the behavior of radionuclides in the environment, radiation
safety specialists, etc. The overall workforce, from the operation phase to site reha-
bilitation and monitoring after closure, must be trained in various aspects of the
431Uranium mine and mill remediation and reclamation
additional risks arising from the presence of elevated concentrations of radioactiv-
ity, particularly in areas of health and safety awareness with qualified staff commit-
ted to ongoing knowledge transfer.
16.6 Funding for rehabilitation
It is unlikely that the presence of typical concentrations of radioactivity in uranium
ore and in waste products will give rise to significantly higher mining costs than
that required in other mining industries, as the same good practices are also
demanded of the broader industry (CRC, 2008). Any proposed expenditure is care-
fully balanced against the likely benefit from such measures.
Some extra costs do arise for the management of radioactivity but most of the
rehabilitation costs are required in any case for the management of other potential
pollutants. The public perception of the uranium mining industry is such that good
practice measures must be strictly adhered to and the practices be closely scruti-
nized. Forecast environmental (including radiological) impact must be compared
before and after implementation of any measure. The objective is to minimize
future remedial costs being borne by the taxpayer (see European Bank for
Reconstruction and Development fund for rehabilitation of uranium mining legacy
sites) and to minimize the chance of land being “sterile” for reuse
postrehabilitation.
Organizations currently hold considerable liabilities for past practices and their
impacts. The cost of not undertaking remedial work must be taken into account.
The mine plan costs include financial assurance for closure and remediation
(OECD, 2014; CNSC, 2012; Western Australian government, 2013). Monies are
held as a bond. Full accounting and provision of future costs must be incorporated
into the economics of a project to achieve sustainability over the long term.
The potential for a previously operated site to reopen after shutdown for the
extraction of further uranium or other valuable constituents is one reason for
delayed final closure and reclamation of a site.
16.7 Future trends
What are the long-term trends associated with rehabilitation of uranium mine and
milling sites? These may be financial, health and/or those related to land reuse.
16.7.1 Regulations
There is an inescapable link between human health and the health of the environ-
ment. Regulations will change as the knowledge base relating to the impact of
radioactivity on human and nonhuman biota increases. The development of a
framework for radiological protection of the environment by national authorities
432 Uranium for Nuclear Power
has evolved from publications provided by the ICRP reports from 2007 to 2014).
The Australian radiation protection authority (ARPANSA, 2014) and the authorities
of other states involved in the rehabilitation of uranium mining and milling sites
provide best practice guidance for the assessment of the impact of environmental
exposures. Defined dose rate benchmarks are recommended to ensure that detrimen-
tal effects due to ionizing radiation are minimized. This framework is the basis for
a best practice approach to environmental protection and provides models to assess
exposure/dose to and the effect on humans and different types of flora and fauna.
Metal toxicity in mining has generally been accepted as having a greater impact
than radioactivity, although it has been suggested that exposure to metal toxicity
may make an organism more susceptible to radiological exposure, that is, there
may be a synergistic effect of radiation dose and exposure to other agents. In many
circumstances, the broader toxicity impacts can be more significant than radiologi-
cal effects (Johnston et al., 2003). Many of these issues are being further studied.
Radiological effects on biota will be better understood, and risks reduced, as data
are obtained for specific flora and fauna and their dose exposure pathways, for
example, the nature of their habitat.
16.7.2 Practice
A uranium mining and milling plan must have regulatory approval and include an
acceptable closure and remediation plan, including rehabilitation costs.
Radiological impact assessments (ARPANSA, 2014) must be carried out throughout
the project. To ensure that effort and resources are not expended unnecessarily,
these should be addressed as soon as possible within the environmental RMP. The
establishment of a uranium mine and mill with the necessary waste disposal facili-
ties is a “nuclear action” practice with subsequent integral site decommissioning
and rehabilitation tasks. Real, potential, or perceived exposure of the environment
and the associated risks must be addressed. A specific scenario at any time during a
project will include a description of the current practice and the radiological issues
(radioactivity sources, pathways to exposure, physicochemical properties of materi-
als containing radioactivity, affected environment, etc.)—see Fig. 16.7.
The application of an environmental protection framework is recommended at
the conceptual/planning, operational, and rehabilitation stages. This will assist in
identifying the sources and nature of the radioactivity, the potential pathways to
exposure, the relevant affected organisms, etc., as well as allow the assessment of
predicted against actual impacts and determine the level of effort/cost for environ-
mental protection.
Extra monitoring over the longer term may be needed to address local concerns
and to demonstrate transparency of project management procedures.
16.7.3 Research
The following are areas of research that would assist an acceptable rehabilitation of
a uranium mine site. Many of these issues are the subject of papers presented at
433Uranium mine and mill remediation and reclamation
meetings such as the IAEA-UMREG meeting in Bad Schlema, Germany, in August
2015. Much of the information from research to reduce the detrimental impacts of a
mine site and to assess the effectiveness of rehabilitation is site specific. It may
include the following topics:
● Up front beneficiation of uranium ore (crushing and milling, ore sorting, flotation, etc.) to
reduce the quantity of finely milled material reporting as chemically contaminated waste
tailings after mining and milling.● The long-term stability of waste storage facilities remains of major concern. Emphasis is
being placed on the development of innovative methods and techniques to increase settled
densities of tailings to improve the integrity of waste facility structures and the biochemi-
cal and geochemical behavior of sealants to reduce the permeability and leachability of
pollutants from rehabilitated areas.● ISL practice is very site dependent and a thorough understanding of the site geology and
hydrology is essential before approval to develop can be given.● The effect of the migration of pollutants from uranium mining and milling waste on biota
requires increased data to assess impact and to reduce the longer-term risk from its impact.● Better data and improved, validated models are required. This includes a better understanding
of the long-term migration of pollutants and the adsorption of radionuclides onto soils.
16.8 Conclusions
An improved understanding of the environmental impact of uranium mining and
milling is required to address concerns that exist for the rehabilitation of sites used
for uranium mining and milling. Waste management, for example, is a major source
of risk in site rehabilitation from both a public perception and scientific viewpoint,
as uncertainties still exist, such as:
● The effect of time and possibly a change in local site conditions on the rehabilitation
methods proposed
Natural background
Environmental transfer
Organisms & pathways
Timescales
Biological endpoints & risk
Source
Figure 16.7 General aspects to consider when building a project scenario.
Source: From ARPANSA (2014).
434 Uranium for Nuclear Power
● Long-term climate change effects● Population density/demography uncertainties/changes● Dose model/regulation changes, for example, linear nonthreshold dose response model
unknowns
The risk is reduced as the effects on biota are better understood as further data are
obtained and improvements are made to models. They are also reduced as waste
disposal methods are adopted that are more practical in a remediation context.
Although ISL and underground mines (with underground disposal of tailings)
have benefits in regard to site reclamation and land reuse, there are still risks asso-
ciated with them. These risks include:
● Some process wastes may not be readily accommodated underground, such as some pro-
cessing reagents, for example, SX reagents, and slimes;● There are risks associated with radon emanation underground and specialized ventilation
and mining techniques are required during operation to accommodate this. The longer-
term benefits of an underground operation may outweigh this risk and allow underground
mining, which will have a significant effect on the available waste management and site
remediation options; and● Uncontrolled flooding and water contamination are risks in underground operations, ISL
sites and from heap leaching operations. This can have a significant effect on costly water
treatment requirements in the short and long term during rehabilitation.
Future investment in knowledge gaps will reduce the risks particularly in regard
to long-term pollutant migration. It will minimize the need for active maintenance
in the longer term and provide solutions that are economically as well as technically
feasible.
References
ARPANSA, 2005. Code of practice and safety guide on radiation protection and radioactive
waste management in mining and mineral processing, RPS 9.
ARPANSA, 2014. Safety guide for radiation protection of the environment, Safety Guide,
DRAFT. Commonwealth of Australia 2014 as represented by the Australian Radiation
Protection and Nuclear Safety Agency (ARPANSA).
Benson, C.H., Albright, W.H., Roesler, A.C., Abichou, T., 2002. Evaluation of final cover
performance: field data for the alternative cover assessment program (ACAP), Waste
Management WM’02, 24�28 February, Tucson, Arizona, USA, Paper 521.
Bollhofer, A., Doering, C., 2015. Long-Term temporal variability of the radon-222 exhalation
flux from a landform covered by low uranium grade waste rock. J. Environ. Radioact.
127, 88�94.
CNSC, 2012. Uranium mining and milling: the facts of a well-regulated industry. Canadian
Nuclear Safety Commission Fact Sheet, July.
CRC, 2008. Tailings and mine waste ’08. In: 12th International Conference on Tailings and
Mine Waste: Liners, Covers and Barriers, ISBN 978-0-415-48634-7.
Cunningham, K., Kristoff, D., Hovdebo, D., Webster, M., 2015. Post-closure management
of mines in Saskatchewan, AusIMM Bulletin, June, pp. 84�86. Original document
435Uranium mine and mill remediation and reclamation